WorldWideScience

Sample records for ast-1 reactor

  1. Coupling of AST-500 heating reactors with desalination facilities

    International Nuclear Information System (INIS)

    Gureyeva, L.V.; Egorov, V.V.; Podberezniy, V.L.

    1997-01-01

    The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab

  2. Coupling of AST-500 heating reactors with desalination facilities

    Energy Technology Data Exchange (ETDEWEB)

    Gureyeva, L V; Egorov, V V [OKBM, Nizhny Novgorod (Russian Federation); Podberezniy, V L [Scientific Research Inst. of Machine Building, Ekaterinburg (Russian Federation)

    1997-09-01

    The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab.

  3. Coupling of AST-500 heating reactors with desalination facilities

    International Nuclear Information System (INIS)

    Kourachenkov, A.V.

    1998-01-01

    The general issues regarding NHR and desalination facility joint operation for potable water production are briefly considered. AST-500 reactor plant and DOU GTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. Similarity of NHR operation for a heating grid and a desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author)

  4. Basic design decisions for advanced AST-type NHRs

    International Nuclear Information System (INIS)

    Gureyeva, L.V.; Egorov, V.V.; Malamud, V.A.

    1997-01-01

    On the basis of the AST-500 reference design decisions and of the experience gained in the RF during the pilot NDHPs development and construction, the advanced NHR AST-500M has been developed recently by OKB Mechanical Engineering, as well as a whole series of heating and co-generation reactor plants of various unit power. All the designs represent enhanced safety reactor plants meeting the contemporary national requirements and international recommendations for nuclear plants of the new generation. The main objectives for the advanced NHR development are considered. New design decisions and engineering improvements are described briefly. (author). 3 refs, 4 figs

  5. Basic design decisions for advanced AST-type NHRs

    Energy Technology Data Exchange (ETDEWEB)

    Gureyeva, L V; Egorov, V V; Malamud, V A [OKBM, Nizhny Novgorod (Russian Federation)

    1997-09-01

    On the basis of the AST-500 reference design decisions and of the experience gained in the RF during the pilot NDHPs development and construction, the advanced NHR AST-500M has been developed recently by OKB Mechanical Engineering, as well as a whole series of heating and co-generation reactor plants of various unit power. All the designs represent enhanced safety reactor plants meeting the contemporary national requirements and international recommendations for nuclear plants of the new generation. The main objectives for the advanced NHR development are considered. New design decisions and engineering improvements are described briefly. (author). 3 refs, 4 figs.

  6. Theoretical model for investigating the dynamic behaviour of the AST-500 type nuclear heating station reactor

    International Nuclear Information System (INIS)

    Grundmann, U.; Rohde, U.; Naumann, B.

    1985-01-01

    Studies on theoretical simulation of the dynamic behaviour of the AST-500 type reactor primary coolant system are summarized. The first version of a dynamic model in the form of the DYNAST code is described. The DYNAST code is based on a one-dimensional description of the primary coolant circuit including core, draught stack, and intermediate heat exchanger, a vapour dome model, and the point model of neutron kinetics. With the aid of the steady-state computational part of the DYNAST code, studies have been performed on different steady-state operating conditions. Furthermore, some methodological investigations on generalization and improvement of the dynamic model are considered and results presented. (author)

  7. Safety Assessment for transient event occurred during the ASTS test of Hanbit Unit 2

    International Nuclear Information System (INIS)

    Yang, Changkeun; Kim, Yohan; Ha, Sangjun

    2014-01-01

    Safety Injection has been actuated during the ASTS (Automatic Seismic Trip System) test of Hanbit Unit 2 on Feb. 28, 2014. It could be bad effect on system integrity. KHNP has been performed safety assessment of system for effect of Safety Injection (SI) actuation occurred during the ASTS test of hanbit Unit 2. Stable state of nuclear power plant system has been confirmed according to Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the power plant is in a safe condition were conformed. After ASTS action, thermal elimination has been processed throughout the turbine until turbine signal occurrence because ASTS is connected to M-G set in the present hanbit Unit 2. Therefore, Safety Injection signal has been actuated by rapid reduction of Steam Generator pressure. In this paper, it is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation. Stable state of nuclear power plant system has been confirmed for Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the plant is in a safe condition were conformed. It is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation

  8. Aspartate aminotransferase (AST) blood test

    Science.gov (United States)

    ... gov/ency/article/003472.htm Aspartate aminotransferase (AST) blood test To use the sharing features on this page, please enable JavaScript. The aspartate aminotransferase (AST) blood test measures the level of the enzyme AST in ...

  9. AST-500 safety analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Falikov, A A; Bakhmetiev, A M; Kuul, V S; Samoilov, O B [OKBM, Nizhny Novgorod (Russian Federation)

    1997-09-01

    Characteristic AST-type NHR safety features and requirements are described briefly. The main approaches and results of design and beyond-design accidents analyses for the AST-500 NHR, and the results of probabilistic safety assessments are considered. It is concluded that the AST-500 possesses a high safety level in virtue of the development and realization in the design of self-protection, passivity and defence-in-depth principles. (author). 9 refs, 2 figs.

  10. Experimental study of steam bubble velocities and dimensions in the draught trunk of the AST-500 reactor simulator

    International Nuclear Information System (INIS)

    Shanin, V.K.; Drobkov, V.P.; Kulakov, I.V.; Khalmeh, M.V.

    1988-01-01

    Local characteristics for two-phase steam water flow in the vertical channel with 0.45 m diameter and 2 m length, which is the draught trunk of the AST-500 reactor simulator, are investigated. Steam bubble velocities and dimensions were determined by the time-of-flight method using the twinned conductometric transducers. The data obtained testify to the existance of unstable circulation flows in the trunk peripheral region. These flows effect considerably the steam phase motion in the trunk middle part. At the same time the circulation flows to a lesser degree affect steam bubble motion in the trunk low peripheral part and to the lesser degree affect the steam phase in the axial zone near the outlet from the heating section. So the data obtained confirm the conclusion, made earlier, about steam-water flow acceleration in the draught trunk central part

  11. Functional analysis of the ASTE11 gene from the dimorphic yeast Arxula adeninivorans

    International Nuclear Information System (INIS)

    El Fiki, A.; El Metabteb, G.; Boer, E.; Kunze, G.

    2010-01-01

    Arxula adeninivorans is dimorphic yeast with unusual biochemical and physiological characteristic. It is thermo- and osmo- resistance and it can use a wide range of carbon sources for growth. One kinase of the HOG pathway, the MAPKKK is encoded by ASTE11 gene which was isolated from A. adeninivorans. The aste11 mutant was achieved by gene disruption procedure. The Sck1p gene encoding MAPKKK in S. cerevisiae can complement with aste11 mutation. Growth rate of G1211/pAL-ALEU2m, G1211/pAL-ALEU2m-ASTE11 (over-expression transformants) and IS1 [aleu2 aste11 ALEU2] (aste11 mutant), the ASTE11 expression level dose not correlates with salt resistance. However, the growth rate of G1211/pAL-ALEU2m, G1211/pAL-ALEU2m-ASTE11 (over-expression transformants) and IS1 [aleu2 aste11::ALEU2] (aste11 mutant) and the response to thermo stress were affected in the deleted mutant, the Aste11p influenced the thermo resistance of A. adeninivorans. The MAPKKK encoding by STE11 gene from various yeast species is involved in the mating process. The mutant strains and their transformants were lost the capacity to mate. Assessment of the ASTE11 promoter activity with lacZ reporter gene confirmed its inducibility by osmolaytes.

  12. AST Test: MedlinePlus Lab Test Information

    Science.gov (United States)

    ... page: https://medlineplus.gov/labtests/asttest.html AST Test To use the sharing features on this page, please enable JavaScript. What is an AST Test? AST (aspartate aminotransferase) is an enzyme that is ...

  13. Effect of AST-120 on Endothelial Dysfunction in Adenine-Induced Uremic Rats

    Directory of Open Access Journals (Sweden)

    Yuko Inami

    2014-01-01

    Full Text Available Aim. Chronic kidney disease (CKD represents endothelial dysfunction. Monocyte adhesion is recognized as the initial step of arteriosclerosis. Indoxyl sulfate (IS is considered to be a risk factor for arteriosclerosis in CKD. Oral adsorbent AST-120 retards deterioration of renal function, reducing accumulation of IS. In the present study, we determined the monocyte adhesion in the adenine-induced uremic rats in vivo and effects of AST-120 on the adhesion molecules. Methods. Twenty-four rats were divided into control, control+AST-120, adenine, and adenine+AST-120 groups. The number of monocytes adherent to the endothelium of thoracic aorta by imaging the entire endothelial surface and the mRNA expressions of adhesion and atherosclerosis-related molecules were examined on day 49. The mRNA expressions of ICAM-1 and VCAM-1 in human umbilical vein endothelial cells were also examined. Results. Adenine increased the number of adherent monocytes, and AST-120 suppressed the increase. The monocyte adhesion was related to serum creatinine and IS in sera. Overexpression of VCAM-1 and TGF-β1 mRNA in the arterial walls was observed in uremic rats. IS induced increase of the ICAM-1 and VCAM-1 mRNA expressions in vitro. Conclusion. It appears that uremic condition introduces the monocyte adhesion to arterial wall and AST-120 might inhibit increasing of the monocyte adherence with CKD progression.

  14. PhAST: pharmacophore alignment search tool.

    Science.gov (United States)

    Hähnke, Volker; Hofmann, Bettina; Grgat, Tomislav; Proschak, Ewgenij; Steinhilber, Dieter; Schneider, Gisbert

    2009-04-15

    We present a ligand-based virtual screening technique (PhAST) for rapid hit and lead structure searching in large compound databases. Molecules are represented as strings encoding the distribution of pharmacophoric features on the molecular graph. In contrast to other text-based methods using SMILES strings, we introduce a new form of text representation that describes the pharmacophore of molecules. This string representation opens the opportunity for revealing functional similarity between molecules by sequence alignment techniques in analogy to homology searching in protein or nucleic acid sequence databases. We favorably compared PhAST with other current ligand-based virtual screening methods in a retrospective analysis using the BEDROC metric. In a prospective application, PhAST identified two novel inhibitors of 5-lipoxygenase product formation with minimal experimental effort. This outcome demonstrates the applicability of PhAST to drug discovery projects and provides an innovative concept of sequence-based compound screening with substantial scaffold hopping potential. 2008 Wiley Periodicals, Inc.

  15. AstRoMap European Astrobiology Roadmap.

    Science.gov (United States)

    Horneck, Gerda; Walter, Nicolas; Westall, Frances; Grenfell, John Lee; Martin, William F; Gomez, Felipe; Leuko, Stefan; Lee, Natuschka; Onofri, Silvano; Tsiganis, Kleomenis; Saladino, Raffaele; Pilat-Lohinger, Elke; Palomba, Ernesto; Harrison, Jesse; Rull, Fernando; Muller, Christian; Strazzulla, Giovanni; Brucato, John R; Rettberg, Petra; Capria, Maria Teresa

    2016-03-01

    The European AstRoMap project (supported by the European Commission Seventh Framework Programme) surveyed the state of the art of astrobiology in Europe and beyond and produced the first European roadmap for astrobiology research. In the context of this roadmap, astrobiology is understood as the study of the origin, evolution, and distribution of life in the context of cosmic evolution; this includes habitability in the Solar System and beyond. The AstRoMap Roadmap identifies five research topics, specifies several key scientific objectives for each topic, and suggests ways to achieve all the objectives. The five AstRoMap Research Topics are • Research Topic 1: Origin and Evolution of Planetary Systems • Research Topic 2: Origins of Organic Compounds in Space • Research Topic 3: Rock-Water-Carbon Interactions, Organic Synthesis on Earth, and Steps to Life • Research Topic 4: Life and Habitability • Research Topic 5: Biosignatures as Facilitating Life Detection It is strongly recommended that steps be taken towards the definition and implementation of a European Astrobiology Platform (or Institute) to streamline and optimize the scientific return by using a coordinated infrastructure and funding system.

  16. Asteroids Dynamic Site-AstDyS

    Science.gov (United States)

    Knezevic, Zoran; Milani, Andrea

    2012-08-01

    The AstDyS online information service (http://hamilton.dm.unipi.it/astdys/) contains data on numbered and multi - opposition asteroids, including orbital elements, their uncertainty, proper elements, ephemerides with uncertainty, and more. AstDyS also provides additional scientific output computed from the raw observational data. This value added currently includes: more accurate orbits computed with advanced dynamical and observational error model s; their uncertainty, as expressed by the covariance matrix formalism; ephemerides computed on request for each observer, with uncertainty; mean and proper orbital elements (for this output, AstDyS is the primary source worldwide); statistical quality control, providing a rigorous observational error model. All this is available with a sophisticated web interface, providing multiple search functions and online computations as well as complete orbital and residual files. There are several ways in which the A stDyS service could be expanded and improved in the next future, like the explicit classification of asteroids into asteroid families, the classification of resonant asteroids, and an updated self - consistent population model (to be used, e.g., for survey simulations). The IAU Division I endorsed the proposal for AstDyS to become an IAU (permanent) service, which would include the IAU supervision of the AstDyS system, keeping under control the quality of the work and the continuous update under conditions of scientific competition.

  17. Reliability and Validity of Korean Version of Apraxia Screen of TULIA (K-AST).

    Science.gov (United States)

    Kim, Soo Jin; Yang, You-Na; Lee, Jong Won; Lee, Jin-Youn; Jeong, Eunhwa; Kim, Bo-Ram; Lee, Jongmin

    2016-10-01

    To evaluate the reliability and validity of Korean version of AST (K-AST) as a bedside screening test of apraxia in patients with stroke for early and reliable detection. AST was translated into Korean, and the translated version received authorization from the author of AST. The performances of K-AST in 26 patients (21 males, 5 females; mean age 65.42±17.31 years) with stroke (23 ischemic, 3 hemorrhagic) were videotaped. To test the reliability and validity of K-AST, the recorded performances were assessed by two physiatrists and two occupational therapists twice at a 1-week interval. The patient performances at admission in Korean version of Mini-Mental State Examination (K-MMSE), self-care and transfer categories of Functional Independence Measure (FIM), and motor praxis area of Loewenstein Occupational Therapy Cognitive Assessment, the second edition (LOTCA-II) were also evaluated. Scores of motor praxis area of LOTCA-II was used to assess the validity of K-AST. Inter-rater reliabilities were 0.983 (preliable and valid test for bedside screening of apraxia.

  18. Variable Stars Observed in the Galactic Disk by AST3-1 from Dome A, Antarctica

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Lingzhi; Ma, Bin; Hu, Yi; Liu, Qiang; Shang, Zhaohui [Key Laboratory of Optical Astronomy, National Astronomical Observatories, Chinese Academy of Sciences, Beijing 100012 (China); Li, Gang; Fu, Jianning [Department of Astronomy, Beijing Normal University, Beijing, 100875 (China); Wang, Lifan; Cui, Xiangqun; Du, Fujia; Gong, Xuefei; Li, Xiaoyan; Li, Zhengyang; Yuan, Xiangyan; Zhou, Jilin [Chinese Center for Antarctic Astronomy, Nanjing 210008 (China); Ashley, Michael C. B. [School of Physics, University of New South Wales, NSW 2052 (Australia); Pennypacker, Carl R. [Center for Astrophysics, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); York, Donald G., E-mail: wanglingzhi@bao.ac.cn [Department of Astronomy and Astrophysics and Enrico Fermi Institute, University of Chicago, Chicago, IL 60637 (United States)

    2017-03-01

    AST3-1 is the second-generation wide-field optical photometric telescope dedicated to time-domain astronomy at Dome A, Antarctica. Here, we present the results of an i -band images survey from AST3-1 toward one Galactic disk field. Based on time-series photometry of 92,583 stars, 560 variable stars were detected with i magnitude ≤16.5 mag during eight days of observations; 339 of these are previously unknown variables. We tentatively classify the 560 variables as 285 eclipsing binaries (EW, EB, and EA), 27 pulsating variable stars ( δ Scuti, γ Doradus, δ Cephei variable, and RR Lyrae stars), and 248 other types of variables (unclassified periodic, multiperiodic, and aperiodic variable stars). Of the eclipsing binaries, 34 show O’Connell effects. One of the aperiodic variables shows a plateau light curve and another variable shows a secondary maximum after peak brightness. We also detected a complex binary system with an RS CVn-like light-curve morphology; this object is being followed-up spectroscopically using the Gemini South telescope.

  19. Pattern of AST and ALT changes in Relation to Hemolysis in sickle cell Disease

    Directory of Open Access Journals (Sweden)

    K. Nsiah

    2011-01-01

    Full Text Available Background Elevated aminotransferase levels are commonly associated with compromised hepatic integrity from various insults. In sickle cell disease, aspartate transaminase (AST is also released via intravascular hemolysis. This study was done to determine the pattern of changes in AST and alanine transaminase (ALT, in particular the AST:ALT ratio, and to relate these to the hemolytic state, which we consider to be more important than hepatic and cardiac dysfunction in some individuals with sickle cell disease. Methods Serum aminotransferase levels were measured in 330 subjects with sickle cell disease, as well as hemoglobin, reticulocytes, and lactate dehydrogenase. The AST:ALT ratio was designated as a hemolytic marker, and simple and multivariate regression analyses were carried out between this ratio and other hemolytic markers. Results Mean AST and ALT levels were 48.24 % 27.78 and 26.48 % 22.73 U/L, respectively. However, for 49 subjects without sickle cell disease, mean AST and ALT levels were the same, ie, 23.0 U/L. In the subjects with sickle cell disease, the increases in AST levels were far higher than for ALT, supporting its release via intravascular hemolysis. In 95.8% of the subjects with sickle cell disease, the AST:ALT ratio was > 1, but our results did not suggest overt malfunctioning of the liver and heart in the majority of subjects. Conclusion Regression analyses support the use of the AST:ALT ratio as a hemolytic marker, because it has an inverse association with the hemoglobin level. Whether in steady state or in crisis, provided hepatic and cardiac integrity has not been compromised, subjects with sickle cell disease would have higher AST levels due to the hemolytic nature of the condition. This is the first report highlighting the AST:ALT ratio in sickle cell disease.

  20. ELEVATED ALT AND AST IN AN ASYMPTOMATIC PERSON

    Directory of Open Access Journals (Sweden)

    KEW ST

    2009-01-01

    Full Text Available -Abnormal liver function test with raised alanine aminotransferase (ALT and raised aspartate aminotransferase (AST are commonly seen in primary care setting. -Chronic alcohol consumption, drugs, non-alcoholic steatohepatitis (NASH and chronic viral hepatitis are common causes associated with raised ALT and AST. -In chronic viral hepatitis, the elevation of liver enzyme may not correlate well with the degree of liver damage. -Non-hepatic causes of raised ALT and AST include polymyositis, acute muscles injury, acute myocardial infarction and hypothyroidism. -In the primary care setting, the doctor should obtain a complete history regarding the risk factors for viral hepatitis, substance abuse and request investigations accordingly. -Suspected chronic viral hepatitis and liver cirrhosis are best referred to hepatologist for further management.

  1. Astakine LvAST binds to the β subunit of F1-ATP synthase and likely plays a role in white shrimp Litopeneaus vannamei defense against white spot syndrome virus.

    Science.gov (United States)

    Liang, Gao-Feng; Liang, Yan; Xue, Qinggang; Lu, Jin-Feng; Cheng, Jun-Jun; Huang, Jie

    2015-03-01

    Cytokines play a critical role in innate and adaptive immunity. Astakines represent a group of invertebrate cytokines that are related to vertebrate prokineticin and function in promoting hematopoiesis in crustaceans. We have identified an astakine from the white shrimp Litopeneaus vannamei and named it LvAST in a previous research. In the present research, we investigated the interactions among LvAST, the envelope protein VP37 of white spot syndrome virus (i.e., WSSV), and the β subunit of F1-ATP synthase (ATPsyn-β) of the white shrimp (i.e., BP53) using binding assays and co-precipitations. We also examined the effects of LvAST on shrimp susceptibility to WSSV. We found that LvAST and VP37 competitively bound to BP53, but did not bind to each other. Shrimps that had been injected with recombinant LvAST exhibited significantly lower mortality and longer survival time in experimental infections by WSSV. In contrast, shrimps whose LvAST gene expression had been inhibited by RNA interference showed significantly higher WSSV infection intensity and shorter survival time following viral challenges. These results suggested that LvAST and WSSV both likely use ATPsyn-β as a receptor and LvAST plays a role in shrimp defense against WSSV infection. This represented the first research showing the involvement of astakines in host antiviral immunity. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. A new method for synthesis of As-Te chalcogenide films

    Science.gov (United States)

    Mochalov, Leonid; Nezhdanov, Aleksey; Usanov, Dmitry; Markelov, Aleksey; Trushin, Vladimir; Chidichimo, Giuseppe; De Filpo, Giovanni; Gogova, Daniela; Mashin, Aleksandr

    2017-11-01

    A novel Plasma Enhanced Chemical Vapor Deposition method for synthesis of amorphous AsxTe100-x (31 ≤ x ≤ 49) films is demonstrated. The innovative process has been developed in a non-equilibrium low-temperature argon plasma under reduced pressure, employing for the first time volatile As and Te as precursors. Utilization of inorganic precursors, in contrast to the typically used in CVD metal-organic precursors, has given us the chance to achieve ≿halcogenide As-Te films of very high quality and purity. Phase and structural evolution of the As-Te system, based on equilibrium coexistence of two phases (AsTe and As2Te3) has been studied. The dependence of structure and optical bandgap of the chalcogenide materials on their composition was established. The newly developed process is cost-effective and enables deposition of As-Te films with a thickness ranging from 10 nm to 10 μm, the latter is highly desireable for one-mode planar waveguides applications and in other components of integral optics.

  3. AST1306, a novel irreversible inhibitor of the epidermal growth factor receptor 1 and 2, exhibits antitumor activity both in vitro and in vivo.

    Directory of Open Access Journals (Sweden)

    Hua Xie

    Full Text Available Despite the initial response to the reversible, ATP-competitive quinazoline inhibitors that target ErbB-family, such a subset of cancer patients almost invariably develop resistance. Recent studies have provided compelling evidence that irreversible ErbB inhibitors have the potential to override this resistance. Here, we found that AST1306, a novel anilino-quinazoline compound, inhibited the enzymatic activities of wild-type epidermal growth factor receptor (EGFR and ErbB2 as well as EGFR resistant mutant in both cell-free and cell-based systems. Importantly, AST1306 functions as an irreversible inhibitor, most likely through covalent interaction with Cys797 and Cys805 in the catalytic domains of EGFR and ErbB2, respectively. Further studies showed that AST1306 inactivated pathways downstream of these receptors and thereby inhibited the proliferation of a panel of cancer cell lines. Although the activities of EGFR and ErbB2 were similarly sensitive to AST1306, ErbB2-overexpressing cell lines consistently exhibited more sensitivity to AST1306 antiproliferative effects. Consistent with this, knockdown of ErbB2, but not EGFR, decreased the sensitivity of SK-OV-3 cells to AST1306. In vivo, AST1306 potently suppressed tumor growth in ErbB2-overexpressing adenocarcinoma xenograft and FVB-2/N(neu transgenic breast cancer mouse models, but weakly inhibited the growth of EGFR-overexpressing tumor xenografts. Tumor growth inhibition induced by a single dose of AST1306 in the SK-OV-3 xenograft model was accompanied by a rapid (within 2 h and sustained (≥24 h inhibition of both EGFR and ErbB2, consistent with an irreversible inhibition mechanism. Taken together, these results establish AST1306 as a selective, irreversible ErbB2 and EGFR inhibitor whose growth-inhibitory effects are more potent in ErbB2-overexpressing cells.

  4. Kirjavahetus Kaarel Robert Pustaga / Karl Ast Rumor

    Index Scriptorium Estoniae

    Rumor, Karl, pseud., 1886-1971

    2009-01-01

    Kirjavahetus Rio de Janeiros elava Karl Ast Rumori ja New Yorgis elava Kaarel Robert Pusta vahel Eesti kongressi kokkukutsumise vajalikkuse, väliseesti ühingute ja nõukogude, ajakirjanduse ning üldise poliitilise olukorra teemadel

  5. Translation and cultural adaptation of the Aguado Syntax Test (AST) into Brazilian Portuguese.

    Science.gov (United States)

    Baggio, Gustavo Inheta; Hage, Simone Rocha de Vasconcellos

    2017-12-07

    To perform the translation and cultural adaptation of the Aguado Syntax Test (AST) into Brazilian Portuguese considering the linguistic and cultural reality of the language. The AST assesses the early morphosyntactic development in children aged 3 to 7 in terms of understanding and expression of various types of structures such as sentences, pronouns, verbal voices, comparisons, prepositions and verbal desinence as to number, mode and tense. The process of translation and cultural adaptation followed four steps: 1) preparation of two translations; 2) synthesis of consensual translations; 3) backtranslation; and 4) verification of equivalence between the initial translations and backtranslations that resulted in the final translated version. The whole process of translation and cultural adaptation revealed the presence of equivalence and reconciliation of the translated items and an almost complete semantic equivalence between the two translations and the absence of consistent translation difficulties. The AST was translated and culturally adapted into Brazilian Portuguese, constituting the first step towards validation and standardization of the test.

  6. AGS Spallation Target Experiment (ASTE) Collaboration

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1999-01-01

    An experiment on mercury spallation target with high energy proton beam, called as the AGS Spallation Target Experiment (ASTE) Collaboration, has been performed at Alternating Gradient Synchrotron (AGS) of Brookhaven National Laboratory (BNL) in USA, in cooperation among the laboratories in Japan, Europe and USA. The experimental setup, scope and preliminary results are presented in the paper. (author)

  7. Biodegradation of Chlorpyrifos by Pseudomonas Resinovarans Strain AST2.2 Isolated from Enriched Cultures.

    OpenAIRE

    Anish Sharma*,; Jyotsana Pandit; Ruchika Sharma and; Poonam Shirkot

    2016-01-01

    A bacterial strain AST2.2 with chlorpyrifos degrading ability was isolated by enrichment technique from apple orchard soil with previous history of chlorpyrifos use. Based on the morphological, biochemical tests and 16S rRNA sequence analysis, AST2.2 strain was identified as Pseudomonas resinovarans. The strain AST2.2 utilized chlorpyrifos as the sole source of carbon and energy. This strain exhibited growth upto 400mg/l concentration of chlorpyrifos and exhibited high extracellular organopho...

  8. Effect of an Oral Adsorbent, AST-120, on Dialysis Initiation and Survival in Patients with Chronic Kidney Disease

    Directory of Open Access Journals (Sweden)

    Shingo Hatakeyama

    2012-01-01

    Full Text Available The oral adsorbent AST-120 has the potential to delay dialysis initiation and improve survival of patients on dialysis. We evaluated the effect of AST-120 on dialysis initiation and its potential to improve survival in patients with chronic kidney disease. The present retrospective pair-matched study included 560 patients, grouped according to whether or not they received AST-120 before dialysis (AST-120 and non-AST-120 groups. The cumulative dialysis initiation free rate and survival rate were compared by the Kaplan-Meier method. Multivariate analysis was used to determine the impact of AST-120 on dialysis initiation. Our results showed significant differences in the 12- and 24-month dialysis initiation free rate (P<0.001, although no significant difference was observed in the survival rate between the two groups. In conclusion, AST-120 delays dialysis initiation in chronic kidney disease (CKD patients but has no effect on survival. AST-120 is an effective therapy for delaying the progression of CKD.

  9. Precision holography for N={2}^{\\ast } on S 4 from type IIB supergravity

    Science.gov (United States)

    Bobev, Nikolay; Gautason, Friðrik Freyr; van Muiden, Jesse

    2018-04-01

    We find a new supersymmetric solution of type IIB supergravity which is holographically dual to the planar limit of the four-dimensional N={2}^{\\ast } supersymmetric Yang-Mills theory on S 4. We study a probe fundamental string in this background which is dual to a supersymmetric Wilson loop in the N={2}^{\\ast } theory. Using holography we calculate the expectation value of this line operator to leading order in the 't Hooft coupling. The result is a non-trivial function of the mass parameter of the N={2}^{\\ast } theory that precisely matches the result from supersymmetric localization.

  10. Dupuytren’s and Ledderhose Diseases in a Family with LMNA-Related Cardiomyopathy and a Novel Variant in the ASTE1 Gene

    Directory of Open Access Journals (Sweden)

    Michael V. Zaragoza

    2017-11-01

    Full Text Available Dupuytren’s disease (palmar fibromatosis involves nodules in fascia of the hand that leads to flexion contractures. Ledderhose disease (plantar fibromatosis is similar with nodules of the foot. While clinical aspects are well-described, genetic mechanisms are unknown. We report a family with cardiac disease due to a heterozygous LMNA mutation (c.736C>T, p.Gln246Stop with palmar/plantar fibromatosis and investigate the hypothesis that a second rare DNA variant increases the risk for fibrotic disease in LMNA mutation carriers. The proband and six family members were evaluated for the cardiac and hand/feet phenotypes and tested for the LMNA mutation. Fibroblast RNA studies revealed monoallelic expression of the normal LMNA allele and reduced lamin A/C mRNAs consistent with LMNA haploinsufficiency. A novel, heterozygous missense variant (c.230T>C, p.Val77Ala in the Asteroid Homolog 1 (ASTE1 gene was identified as a potential risk factor in fibrotic disease using exome sequencing and family studies of five family members: four LMNA mutation carriers with fibromatosis and one individual without the LMNA mutation and no fibromatosis. With a possible role in epidermal growth factor receptor signaling, ASTE1 may contribute to the increased risk for palmar/plantar fibromatosis in patients with Lamin A/C haploinsufficiency.

  11. LHCb: Measurement of the polarization amplitudes of the decay $B^0 \\rightarrow J/\\psi K^\\ast$

    CERN Multimedia

    Linn, C

    2011-01-01

    Using the data sample recorded with the LHCb detector in 2010 we perform a combined angular and lifetime analysis of the decay $B^0 \\rightarrow J/\\psi K^\\ast$. The data corresponds to an integrated luminosity of about 36 pb$^{-1}$ and was taken at the LHC at an centre-of-mass energy of $\\sqrt{s}$= 7 TeV. A total of 3909 $J/\\psi K^*$ candidates are found and are used to extract the polarisation amplitudes and the corresponding strong phases for the decays $B_d \\rightarrow J/\\psi K^\\ast$.

  12. Measurement of $H{\\to}W^\\pm W^{\\mp\\ast}{\\to}\\ell^-\\bar{\

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00423318; Köneke, Karsten

    This thesis presents and discusses measurements of the coupling of the Higgs boson to vector bosons, using data collected at $\\sqrt{s}=13$ TeV by the ATLAS detector. A full analysis of the first $5.8$ fb${}^{-1}$ of LHC Run 2 data investigating the $H{\\to}W^\\pm W^{\\mp\\ast}{\\to}\\ell^-\\bar{\

  13. Incremental Predictive Value of Serum AST-to-ALT Ratio for Incident Metabolic Syndrome: The ARIRANG Study

    Science.gov (United States)

    Ahn, Song Vogue; Baik, Soon Koo; Cho, Youn zoo; Koh, Sang Baek; Huh, Ji Hye; Chang, Yoosoo; Sung, Ki-Chul; Kim, Jang Young

    2016-01-01

    Aims The ratio of aspartate aminotransferase (AST) to alanine aminotransferase (ALT) is of great interest as a possible novel marker of metabolic syndrome. However, longitudinal studies emphasizing the incremental predictive value of the AST-to-ALT ratio in diagnosing individuals at higher risk of developing metabolic syndrome are very scarce. Therefore, our study aimed to evaluate the AST-to-ALT ratio as an incremental predictor of new onset metabolic syndrome in a population-based cohort study. Material and Methods The population-based cohort study included 2276 adults (903 men and 1373 women) aged 40–70 years, who participated from 2005–2008 (baseline) without metabolic syndrome and were followed up from 2008–2011. Metabolic syndrome was defined according to the harmonized definition of metabolic syndrome. Serum concentrations of AST and ALT were determined by enzymatic methods. Results During an average follow-up period of 2.6-years, 395 individuals (17.4%) developed metabolic syndrome. In a multivariable adjusted model, the odds ratio (95% confidence interval) for new onset of metabolic syndrome, comparing the fourth quartile to the first quartile of the AST-to-ALT ratio, was 0.598 (0.422–0.853). The AST-to-ALT ratio also improved the area under the receiver operating characteristic curve (AUC) for predicting new cases of metabolic syndrome (0.715 vs. 0.732, P = 0.004). The net reclassification improvement of prediction models including the AST-to-ALT ratio was 0.23 (95% CI: 0.124–0.337, Pmetabolic syndrome and had incremental predictive value for incident metabolic syndrome. PMID:27560931

  14. Correlation between Aminotransferase Ratio (AST/ALT and Other Biochemical Parameters in Chronic Liver Disease of Viral Origin

    Directory of Open Access Journals (Sweden)

    Shah Md Fazlul Karim

    2015-03-01

    Full Text Available Background: In recent years the ratio of aspartate aminotransferase (AST to alanine aminotransferase (ALT in patients of chronic liver disease (CLD of various origins has gained much attention. This variable is readily available, easy to interpret, and inexpensive and the clinical utility of the AST/ALT ratio in the diagnostic workup of patients with CLD is quite promising. Objective: The present study was designed to find out the link between aminotransferase (AST/ALT ratio with commonly measured biochemical parameters of liver function tests in CLD of viral origin. Materials and method: This cross sectional study was carried out in the department of Biochemistry, Sir Salimullah Medical College, Dhaka, Bangladesh. Forty four biopsy proven diagnosed subjects of chronic viral hepatitis without cirrhosis of both sex were selected purposively. With aseptic precaution 5 mL venous blood was collected from each subject and common liver function tests (serum AST, ALT, AST/ALT ratio, alkaline phosphatase, total bilirubin, serum total protein, serum albumin, serum globulin, serum albumin/globulin ratio, prothrombin time and viral serology (HBsAg, Anti HDV antibody, Anti HCV antibody were performed. Data were analyzed by SPSS version 19 for Windows. Pearson’s correlation test was done to determine association between AST/ALT with other biochemical parameters. Results: Mean(±SD age of the study subjects was 32.55±10.55 years (range 20-50 years with 48 (77.7% male and 14 (22.6% female subjects. Pearson’s correlation test was done between AST to ALT ratio with other biochemical parameters and prothrombin time showed significant positive correlation (p <0.01. Conclusion: In our study we found significant positive correlation between AST/ALT with prothrombin time in CLD subjects without cirrhosis.

  15. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  16. Reliability of fast reactor mixed-oxide fuel during operational transients

    International Nuclear Information System (INIS)

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

  17. Influence of Preoperative Serum Aspartate Aminotransferase (AST Level on the Prognosis of Patients with Non-Small Cell Lung Cancer

    Directory of Open Access Journals (Sweden)

    Shu-Lin Chen

    2016-09-01

    Full Text Available The purpose of this work is to analyze preoperative serum aspartate aminotransferase (AST levels and their effect on the prognosis of patients with non-small cell lung cancer (NSCLC after surgical operation. These analyses were performed retrospectively in patients with NSCLC followed by surgery; participants were recruited between January 2004 and January 2008. All clinical information and laboratory results were collected from medical records. We explored the association between preoperative serum AST and recurrence-free survival (RFS, and the overall survival (OS of NSCLC patients. Kaplan–Meier analysis and Cox multivariate analysis, stratified by the AST median value, were used to evaluate the prognostic effect. A chi-squared test was performed to compare clinical characteristics in different subgroups. A p-value of ≤0.05 was considered to be statistically significant. A total of 231 patients were enrolled. The median RFS and OS were 22 and 59 months, respectively. The AST levels were divided into two groups, using a cut-off value of 19 U/L: High AST (>19 U/L, n = 113 vs. low AST (≤19 U/L, n = 118. Multivariate analysis indicated that preoperative serum AST > 19 U/L (hazard ratio (HR = 0.685, 95% confidence interval (CI: 0.493–0.994, p = 0.046 for RFS, HR = 0.646, 95% CI: 0.438–0.954, p = 0.028 for OS was an independent prognostic factor for both RFS and OS. High preoperative serum AST levels may serve as a valuable marker to predict the prognosis of NSCLC after operation.

  18. AzTEC on ASTE Survey of Submillimeter Galaxies

    Science.gov (United States)

    Kohno, K.; Tamura, Y.; Hatsukade, B.; Nakanishi, K.; Iono, D.; Takata, T.; Wilson, G. W.; Yun, M. S.; Perera, T.; Austermann, J. E.; Scott, K. S.; Hughes, H.; Aretxaga, I.; Tanaka, K.; Oshima, T.; Yamaguchi, N.; Matsuo, H.; Ezawa, H.; Kawabe, R.

    2008-10-01

    We have conducted an unprecedented survey of submillimeter galaxies (SMGs) using the 144 pixel bolometer camera AzTEC mounted on the ASTE 10-m dish in Chile. We have already obtained many (>20) wide (typically 12' × 12' or wider) and deep (1 σ sensitivity of 0.5-1.0 mJy) 1.1 mm continuum images of known blank fields and over-density regions/protoclusters across a wide range of redshifts with a spatial resolution of ˜ 30''. It has resulted in the numerous (˜ a few 100, almost equivalent to the total number of the previously known SMGs) new and secure detections of SMGs. In this paper, we present initial results of two selected fields, SSA 22 and AKARI Deep Field South (ADF-S). A significnat clustering of bright SMGs toward the density peak of LAEs is found in SSA 22. We derived the differential and cumulative number counts from the detected sources in ADF-S, which probe the faintest flux densities (down to ˜1 mJy) among 1-mm blank field surveys to date.

  19. X-ray electron spectra of chalcogenide glasses and polycrystalline alloys of Ge-Te and As-Te systems

    International Nuclear Information System (INIS)

    Panus, V.R.

    1990-01-01

    Comparative investigation into structures of crystals and glasses in Ge-Te and As-Te two-component systems was conducted. Analysis of x-ray electron spectra of Ge-Te and As-Te systems indicates, that processes of dissociation-association resulting in formation of new structure units occur in telluride melts at synthesis temperatures. Structural chemical composition of binary glass-like alloys of Ge-Te and As-Te systems differs essentially from the one that corresponds to fusibility equilibrium curve. Oxygen doping into tellurium-base glasses results mainly in occurence of structures forecasted due to thermochemical calculation

  20. Development of a Seismic Setpoint Calculation Methodology Using a Safety System Approach

    International Nuclear Information System (INIS)

    Lee, Chang Jae; Baik, Kwang Il; Lee, Sang Jeong

    2013-01-01

    The Automatic Seismic Trip System (ASTS) automatically actuates reactor trip when it detects seismic activities whose magnitudes are comparable to a Safe Shutdown Earthquake (SSE), which is the maximum hypothetical earthquake at the nuclear power plant site. To ensure that the reactor is tripped before the magnitude of earthquake exceeds the SSE, it is crucial to reasonably determine the seismic setpoint. The trip setpoint and allowable value for the ASTS for Advanced Power Reactor (APR) 1400 Nuclear Power Plants (NPPs) were determined by the methodology presented in this paper. The ASTS that trips the reactor when a large earthquake occurs is categorized as a non safety system because the system is not required by design basis event criteria. This means ASTS has neither specific analytical limit nor dedicated setpoint calculation methodology. Therefore, we developed the ASTS setpoint calculation methodology by conservatively considering that of PPS. By incorporating the developed methodology into the ASTS for APR1400, the more conservative trip setpoint and allowable value were determined. In addition, the ZPA from the Operating Basis Earthquake (OBE) FRS of the floor where the sensor module is located is 0.1g. Thus, the allowance of 0.17g between OBE of 0.1 g and ASTS trip setpoint of 0.27 g is sufficient to prevent the reactor trip before the magnitude of the earthquake exceeds the OBE. In result, the developed ASTS setpoint calculation methodology is evaluated as reasonable in both aspects of the safety and performance of the NPPs. This will be used to determine the ASTS trip setpoint and allowable for newly constructed plants

  1. Non-amidated and amidated members of the C-type allatostatin (AST-C) family are differentially distributed in the stomatogastric nervous system of the American lobster, Homarus americanus.

    Science.gov (United States)

    Christie, Andrew E; Miller, Alexandra; Fernandez, Rebecca; Dickinson, Evyn S; Jordan, Audrey; Kohn, Jessica; Youn, Mina C; Dickinson, Patsy S

    2018-01-13

    The crustacean stomatogastric nervous system (STNS) is a well-known model for investigating neuropeptidergic control of rhythmic behavior. Among the peptides known to modulate the STNS are the C-type allatostatins (AST-Cs). In the lobster, Homarus americanus, three AST-Cs are known. Two of these, pQIRYHQCYFNPISCF (AST-C I) and GNGDGRLYWRCYFNAVSCF (AST-C III), have non-amidated C-termini, while the third, SYWKQCAFNAVSCFamide (AST-C II), is C-terminally amidated. Here, antibodies were generated against one of the non-amidated peptides (AST-C I) and against the amidated isoform (AST-C II). Specificity tests show that the AST-C I antibody cross-reacts with both AST-C I and AST-C III, but not AST-C II; the AST-C II antibody does not cross-react with either non-amidated peptide. Wholemount immunohistochemistry shows that both subclasses (non-amidated and amidated) of AST-C are distributed throughout the lobster STNS. Specifically, the antibody that cross-reacts with the two non-amidated peptides labels neuropil in the CoGs and the stomatogastric ganglion (STG), axons in the superior esophageal (son) and stomatogastric (stn) nerves, and ~ 14 somata in each commissural ganglion (CoG). The AST-C II-specific antibody labels neuropil in the CoGs, STG and at the junction of the sons and stn, axons in the sons and stn, ~ 42 somata in each CoG, and two somata in the STG. Double immunolabeling shows that, except for one soma in each CoG, the non-amidated and amidated peptides are present in distinct sets of neuronal profiles. The differential distributions of the two AST-C subclasses suggest that the two peptide groups are likely to serve different modulatory roles in the lobster STNS.

  2. Measurement prospects for VBF $H{\\rightarrow\\,}WW^{(\\ast)}{\\rightarrow\\,}e\

    CERN Document Server

    The ATLAS collaboration

    2016-01-01

    This note presents the prospects for the ATLAS experiment to observe and measure Vector Boson Fusion (VBF) Higgs-boson production, with the Higgs boson decaying into two $W$ bosons in the High Luminosity LHC environment. The production of two forward jets in association with the Higgs boson, as well as the requirement that the $W$ bosons both decay to leptons, provides a distinctive detector signature. The VBF production process has the second largest Higgs-boson production cross-section at the LHC and can be computed with small theoretical uncertainties. These properties allow for precision measurements in the High Luminosity LHC with 3 $\\textrm{ab}^{-1}$ of data. In addition, measurements of the VBF $H{\\rightarrow\\,}WW^{(\\ast)}{\\rightarrow\\,}e\

  3. Kas Gaddafi tõesti põgenes Liibüast? / Heiki Suurkask

    Index Scriptorium Estoniae

    Suurkask, Heiki, 1972-

    2011-01-01

    Suur autokolonn Liibüa diktaatori Muammar Gaddafi lähikondlaste ja varandusega jõudis Liibüast Nigeri kaudu Burkina Fasosse. USA võimude andmeil Gaddafit kolonnis ei viibi. Burkina Faso valitsus väidab, et nad ei paku Gaddafile varjupaika. Kaart: Gaddafi võimalik põgenemistee

  4. AST: an automated sequence-sampling method for improving the taxonomic diversity of gene phylogenetic trees.

    Science.gov (United States)

    Zhou, Chan; Mao, Fenglou; Yin, Yanbin; Huang, Jinling; Gogarten, Johann Peter; Xu, Ying

    2014-01-01

    A challenge in phylogenetic inference of gene trees is how to properly sample a large pool of homologous sequences to derive a good representative subset of sequences. Such a need arises in various applications, e.g. when (1) accuracy-oriented phylogenetic reconstruction methods may not be able to deal with a large pool of sequences due to their high demand in computing resources; (2) applications analyzing a collection of gene trees may prefer to use trees with fewer operational taxonomic units (OTUs), for instance for the detection of horizontal gene transfer events by identifying phylogenetic conflicts; and (3) the pool of available sequences is biased towards extensively studied species. In the past, the creation of subsamples often relied on manual selection. Here we present an Automated sequence-Sampling method for improving the Taxonomic diversity of gene phylogenetic trees, AST, to obtain representative sequences that maximize the taxonomic diversity of the sampled sequences. To demonstrate the effectiveness of AST, we have tested it to solve four problems, namely, inference of the evolutionary histories of the small ribosomal subunit protein S5 of E. coli, 16 S ribosomal RNAs and glycosyl-transferase gene family 8, and a study of ancient horizontal gene transfers from bacteria to plants. Our results show that the resolution of our computational results is almost as good as that of manual inference by domain experts, hence making the tool generally useful to phylogenetic studies by non-phylogeny specialists. The program is available at http://csbl.bmb.uga.edu/~zhouchan/AST.php.

  5. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    Sklenka, L.; Kropik, M.

    2006-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

  6. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

    2006-01-01

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  7. Quantum-field theories as representations of a single $^\\ast$-algebra

    OpenAIRE

    Raab, Andreas

    2013-01-01

    We show that many well-known quantum field theories emerge as representations of a single $^\\ast$-algebra. These include free quantum field theories in flat and curved space-times, lattice quantum field theories, Wightman quantum field theories, and string theories. We prove that such theories can be approximated on lattices, and we give a rigorous definition of the continuum limit of lattice quantum field theories.

  8. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  9. Annual report on JEN-1 reactor

    International Nuclear Information System (INIS)

    Montes, J.

    1972-01-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  10. Safety operation of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    2001-01-01

    There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

  11. Peak Serum AST Is a Better Predictor of Acute Liver Graft Injury after Liver Transplantation When Adjusted for Donor/Recipient BSA Size Mismatch (ASTi

    Directory of Open Access Journals (Sweden)

    Kyota Fukazawa

    2014-01-01

    Full Text Available Background. Despite the marked advances in the perioperative management of the liver transplant recipient, an assessment of clinically significant graft injury following preservation and reperfusion remains difficult. In this study, we hypothesized that size-adjusted AST could better approximate real AST values and consequently provide a better reflection of the extent of graft damage, with better sensitivity and specificity than current criteria. Methods. We reviewed data on 930 orthotopic liver transplant recipients. Size-adjusted AST (ASTi was calculated by dividing peak AST by our previously reported index for donor-recipient size mismatch, the BSAi. The predictive value of ASTi of primary nonfunction (PNF and graft survival was assessed by receiver operating characteristic curve, logistic regression, Kaplan-Meier survival, and Cox proportional hazard model. Results. Size-adjusted peak AST (ASTi was significantly associated with subsequent occurrence of PNF and graft failure. In our study cohort, the prediction of PNF by the combination of ASTi and PT-INR had a higher sensitivity and specificity compared to current UNOS criteria. Conclusions. We conclude that size-adjusted AST (ASTi is a simple, reproducible, and sensitive marker of clinically significant graft damage.

  12. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    2003-01-01

    Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

  13. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    1983-12-01

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

  14. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  15. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  16. Occupational analysis for the Angra-1 reactor

    International Nuclear Information System (INIS)

    Moraes, A.

    1991-01-01

    Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

  17. Multi-hazard Assessment and Scenario Toolbox (MhAST): A Framework for Analyzing Compounding Effects of Multiple Hazards

    Science.gov (United States)

    Sadegh, M.; Moftakhari, H.; AghaKouchak, A.

    2017-12-01

    Many natural hazards are driven by multiple forcing variables, and concurrence/consecutive extreme events significantly increases risk of infrastructure/system failure. It is a common practice to use univariate analysis based upon a perceived ruling driver to estimate design quantiles and/or return periods of extreme events. A multivariate analysis, however, permits modeling simultaneous occurrence of multiple forcing variables. In this presentation, we introduce the Multi-hazard Assessment and Scenario Toolbox (MhAST) that comprehensively analyzes marginal and joint probability distributions of natural hazards. MhAST also offers a wide range of scenarios of return period and design levels and their likelihoods. Contribution of this study is four-fold: 1. comprehensive analysis of marginal and joint probability of multiple drivers through 17 continuous distributions and 26 copulas, 2. multiple scenario analysis of concurrent extremes based upon the most likely joint occurrence, one ruling variable, and weighted random sampling of joint occurrences with similar exceedance probabilities, 3. weighted average scenario analysis based on a expected event, and 4. uncertainty analysis of the most likely joint occurrence scenario using a Bayesian framework.

  18. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    Vlok, J.W.H.

    2003-01-01

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  19. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Karel, Matejka; Lubomir, Sklenka

    2005-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

  20. Concept and designs of new-generation fast reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.

    1993-01-01

    This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

  1. Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

    2013-01-01

    Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

  2. Treatment of arsenic contaminated water in a laboratory scale up-flow bio-column reactor

    International Nuclear Information System (INIS)

    Mondal, P.; Majumder, C.B.; Mohanty, B.

    2008-01-01

    The present paper describes the observations on the treatment of arsenic contaminated synthetic industrial effluent in a bio-column reactor. Ralstonia eutropha MTCC 2487 has been immobilized on the granular activated carbon (GAC) bed in the column reactor. The synthetic water sample containing As(T) (As(III):As(V) = 1:1), Fe, Mn, Cu and Zn at the initial concentrations of 25, 10, 2, 5, 10 ppm, respectively, was used. Concentrations of all the elements have been found to be reduced below their permissible limits in the treated water. The significant effect of empty bed contact time (EBCT) and bed height on the arsenic removal was observed in the initial stage. However, after some time of operation (approximately 3-4 days) no such effect was observed. Removal of As(III) and As(V) was almost similar after ∼2 days of operation. However, at the initial stage As(V) removal was slightly more than that of As(III). In absence of washing, after ∼4-5 days of operation, the bio-column reactor was observed to act as a GAC column reactor based on physico-chemical adsorption. Like arsenic, the percent removals of Fe, Mn, Cu and Zn also attained minimum after ∼1 day and increased significantly to the optimum value within 3-4 days of operation. Dissolved oxygen (DO) has been found to decrease along with the increasing bed height from the bottom. The pH of the solution in the reactor has increased slightly and oxidation-reduction potential (ORP) has decreased with the time of operation

  3. Estimation of radioactivity in structural materials of ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

  4. Single Point Vulnerability Analysis of Automatic Seismic Trip System

    International Nuclear Information System (INIS)

    Oh, Seo Bin; Chung, Soon Il; Lee, Yong Suk; Choi, Byung Pil

    2016-01-01

    Single Point Vulnerability (SPV) analysis is a process used to identify individual equipment whose failure alone will result in a reactor trip, turbine generator failure, or power reduction of more than 50%. Automatic Seismic Trip System (ASTS) is a newly installed system to ensure the safety of plant when earthquake occurs. Since this system directly shuts down the reactor, the failure or malfunction of its system component can cause a reactor trip more frequently than other systems. Therefore, an SPV analysis of ASTS is necessary to maintain its essential performance. To analyze SPV for ASTS, failure mode and effect analysis (FMEA) and fault tree analysis (FTA) was performed. In this study, FMEA and FTA methods were performed to select SPV equipment of ASTS. D/O, D/I, A/I card, seismic sensor, and trip relay had an effect on the reactor trip but their single failure will not cause reactor trip. In conclusion, ASTS is excluded as SPV. These results can be utilized as the basis data for ways to enhance facility reliability such as design modification and improvement of preventive maintenance procedure

  5. Single Point Vulnerability Analysis of Automatic Seismic Trip System

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Seo Bin; Chung, Soon Il; Lee, Yong Suk [FNC Technology Co., Yongin (Korea, Republic of); Choi, Byung Pil [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Single Point Vulnerability (SPV) analysis is a process used to identify individual equipment whose failure alone will result in a reactor trip, turbine generator failure, or power reduction of more than 50%. Automatic Seismic Trip System (ASTS) is a newly installed system to ensure the safety of plant when earthquake occurs. Since this system directly shuts down the reactor, the failure or malfunction of its system component can cause a reactor trip more frequently than other systems. Therefore, an SPV analysis of ASTS is necessary to maintain its essential performance. To analyze SPV for ASTS, failure mode and effect analysis (FMEA) and fault tree analysis (FTA) was performed. In this study, FMEA and FTA methods were performed to select SPV equipment of ASTS. D/O, D/I, A/I card, seismic sensor, and trip relay had an effect on the reactor trip but their single failure will not cause reactor trip. In conclusion, ASTS is excluded as SPV. These results can be utilized as the basis data for ways to enhance facility reliability such as design modification and improvement of preventive maintenance procedure.

  6. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  7. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  8. Angular analysis of $B^0 \\to K^\\ast(892)^0 \\ell^+ \\ell^-$

    CERN Document Server

    Abdesselam, A.; Adamczyk, K.; Aihara, H.; Al Said, S.; Arinstein, K.; Arita, Y.; Asner, D.M.; Aso, T.; Atmacan, H.; Aulchenko, V.; Aushev, T.; Ayad, R.; Aziz, T.; Babu, V.; Badhrees, I.; Bahinipati, S.; Bakich, A.M.; Bala, A.; Ban, Y.; Bansal, V.; Barberio, E.; Barrett, M.; Bartel, W.; Bay, A.; Bedny, I.; Behera, P.; Belhorn, M.; Belous, K.; Besson, D.; Bhardwaj, V.; Bhuyan, B.; Biswal, J.; Bloomfield, T.; Blyth, S.; Bobrov, A.; Bondar, A.; Bonvicini, G.; Bookwalter, C.; Boulahouache, C.; Bozek, A.; Bračko, M.; Breibeck, F.; Brodzicka, J.; Browder, T.E.; Waheed, E.; Červenkov, D.; Chang, M. -C.; Chang, P.; Chao, Y.; Chekelian, V.; Chen, A.; Chen, K. -F.; Chen, P.; Cheon, B.G.; Chilikin, K.; Chistov, R.; Cho, K.; Chobanova, V.; Choi, S. -K.; Choi, Y.; Cinabro, D.; Crnkovic, J.; Dalseno, J.; Danilov, M.; Dash, N.; Di Carlo, S.; Dingfelder, J.; Doležal, Z.; Dossett, D.; Drásal, Z.; Drutskoy, A.; Dubey, S.; Dutta, D.; Dutta, K.; Eidelman, S.; Epifanov, D.; Esen, S.; Farhat, H.; Fast, J.E.; Feindt, M.; Ferber, T.; Frey, A.; Frost, O.; Fulsom, B.G.; Gaur, V.; Gabyshev, N.; Ganguly, S.; Garmash, A.; Getzkow, D.; Gillard, R.; Giordano, F.; Glattauer, R.; Goh, Y.M.; Goldenzweig, P.; Golob, B.; Greenwald, D.; Grosse Perdekamp, M.; Grygier, J.; Grzymkowska, O.; Guo, H.; Haba, J.; Hamer, P.; Han, Y.L.; Hara, K.; Hara, T.; Hasegawa, Y.; Hasenbusch, J.; Hayasaka, K.; Hayashii, H.; He, X.H.; Heck, M.; Hedges, M.T.; Heffernan, D.; Heider, M.; Heller, A.; Higuchi, T.; Himori, S.; Hirose, S.; Horiguchi, T.; Hoshi, Y.; Hoshina, K.; Hou, W. -S.; Hsiung, Y.B.; Hsu, C. -L.; Huschle, M.; Hyun, H.J.; Igarashi, Y.; Iijima, T.; Imamura, M.; Inami, K.; Inguglia, G.; Ishikawa, A.; Itagaki, K.; Itoh, R.; Iwabuchi, M.; Iwasaki, M.; Iwasaki, Y.; Iwata, S.; Jacobs, W.W.; Jaegle, I.; Jeon, H.B.; Joffe, D.; Jones, M.; Joo, K.K.; Julius, T.; Kakuno, H.; Kang, J.H.; Kang, K.H.; Kapusta, P.; Kataoka, S.U.; Kato, E.; Kato, Y.; Katrenko, P.; Kawai, H.; Kawasaki, T.; Keck, T.; Kichimi, H.; Kiesling, C.; Kim, B.H.; Kim, D.Y.; Kim, H.J.; Kim, H. -J.; Kim, J.B.; Kim, J.H.; Kim, K.T.; Kim, M.J.; Kim, S.H.; Kim, S.K.; Kim, Y.J.; Kinoshita, K.; Kleinwort, C.; Klucar, J.; Ko, B.R.; Kobayashi, N.; Koblitz, S.; Kodyš, P.; Koga, Y.; Korpar, S.; Kotchetkov, D.; Kouzes, R.T.; Križan, P.; Krokovny, P.; Kronenbitter, B.; Kuhr, T.; Kumar, R.; Kumita, T.; Kurihara, E.; Kuroki, Y.; Kuzmin, A.; Kvasnička, P.; Kwon, Y. -J.; Lai, Y. -T.; Lange, J.S.; Lee, D.H.; Lee, I.S.; Lee, S. -H.; Leitgab, M.; Leitner, R.; Levit, D.; Lewis, P.; Li, C.H.; Li, H.; Li, J.; Li, L.; Li, X.; Li, Y.; Li Gioi, L.; Libby, J.; Limosani, A.; Liu, C.; Liu, Y.; Liu, Z.Q.; Liventsev, D.; Loos, A.; Louvot, R.; Lubej, M.; Lukin, P.; Luo, T.; MacNaughton, J.; Masuda, M.; Matsuda, T.; Matvienko, D.; Matyja, A.; McOnie, S.; Mikami, Y.; Miyabayashi, K.; Miyachi, Y.; Miyake, H.; Miyata, H.; Miyazaki, Y.; Mizuk, R.; Mohanty, G.B.; Mohanty, S.; Mohapatra, D.; Moll, A.; Moon, H.K.; Mori, T.; Morii, T.; Moser, H. -G.; Müller, T.; Muramatsu, N.; Mussa, R.; Nagamine, T.; Nagasaka, Y.; Nakahama, Y.; Nakamura, I.; Nakamura, K.R.; Nakano, E.; Nakano, H.; Nakano, T.; Nakao, M.; Nakayama, H.; Nakazawa, H.; Nanut, T.; Nath, K.J.; Natkaniec, Z.; Nayak, M.; Nedelkovska, E.; Negishi, K.; Neichi, K.; Ng, C.; Niebuhr, C.; Niiyama, M.; Nisar, N.K.; Nishida, S.; Nishimura, K.; Nitoh, O.; Nozaki, T.; Ogawa, A.; Ogawa, S.; Ohshima, T.; Okuno, S.; Olsen, S.L.; Ono, Y.; Onuki, Y.; Ostrowicz, W.; Oswald, C.; Ozaki, H.; Pakhlov, P.; Pakhlova, G.; Pal, B.; Palka, H.; Panzenböck, E.; Park, C. -S.; Park, C.W.; Park, H.; Park, K.S.; Paul, S.; Peak, L.S.; Pedlar, T.K.; Peng, T.; Pesántez, L.; Pestotnik, R.; Peters, M.; Petrič, M.; Piilonen, L.E.; Poluektov, A.; Prasanth, K.; Prim, M.; Prothmann, K.; Pulvermacher, C.; Purohit, M.V.; Rauch, J.; Reisert, B.; Ribežl, E.; Ritter, M.; Röhrken, M.; Rorie, J.; Rostomyan, A.; Rozanska, M.; Rummel, S.; Ryu, S.; Sahoo, H.; Saito, T.; Sakai, K.; Sakai, Y.; Sandilya, S.; Santel, D.; Santelj, L.; Sanuki, T.; Sasao, N.; Sato, Y.; Savinov, V.; Schlüter, T.; Schneider, O.; Schnell, G.; Schönmeier, P.; Schram, M.; Schwanda, C.; Schwartz, A.J.; Schwenker, B.; Seidl, R.; Seino, Y.; Semmler, D.; Senyo, K.; Seon, O.; Seong, I.S.; Sevior, M.E.; Shang, L.; Shapkin, M.; Shebalin, V.; Shen, C.P.; Shibata, T. -A.; Shibuya, H.; Shinomiya, S.; Shiu, J. -G.; Shwartz, B.; Sibidanov, A.; Simon, F.; Singh, J.B.; Sinha, R.; Smerkol, P.; Sohn, Y. -S.; Sokolov, A.; Soloviev, Y.; Solovieva, E.; Stanič, S.; Starič, M.; Steder, M.; Strube, J.F.; Stypula, J.; Sugihara, S.; Sugiyama, A.; Sumihama, M.; Sumisawa, K.; Sumiyoshi, T.; Suzuki, K.; Suzuki, S.; Suzuki, S.Y.; Suzuki, Z.; Takeichi, H.; Takizawa, M.; Tamponi, U.; Tanaka, M.; Tanaka, S.; Tanida, K.; Taniguchi, N.; Taylor, G.N.; Tenchini, F.; Teramoto, Y.; Tikhomirov, I.; Trabelsi, K.; Trusov, V.; Tse, Y.F.; Tsuboyama, T.; Uchida, M.; Uchida, T.; Uehara, S.; Ueno, K.; Uglov, T.; Unno, Y.; Uno, S.; Uozumi, S.; Urquijo, P.; Ushiroda, Y.; Usov, Y.; Vahsen, S.E.; Van Hulse, C.; Vanhoefer, P.; Varner, G.; Varvell, K.E.; Vervink, K.; Vinokurova, A.; Vorobyev, V.; Vossen, A.; Wagner, M.N.; Wang, C.H.; Wang, J.; Wang, M. -Z.; Wang, P.; Wang, X.L.; Watanabe, M.; Watanabe, Y.; Wedd, R.; Wehle, S.; White, E.; Wiechczynski, J.; Williams, K.M.; Won, E.; Yabsley, B.D.; Yamada, S.; Yamamoto, H.; Yamaoka, J.; Yamashita, Y.; Yamauchi, M.; Yashchenko, S.; Ye, H.; Yelton, J.; Yook, Y.; Yuan, C.Z.; Yusa, Y.; Zhang, C.C.; Zhang, L.M.; Zhang, Z.P.; Zhao, L.; Zhilich, V.; Zhukova, V.; Zhulanov, V.; Ziegler, M.; Zivko, T.; Zupanc, A.; Zwahlen, N.; Zyukova, O.

    2016-01-01

    We present a measurement of angular observables, $P_4'$, $P_5'$, $P_6'$, $P_8'$, in the decay $B^0 \\to K^\\ast(892)^0 \\ell^+ \\ell^-$, where $\\ell^+\\ell^-$ is either $e^+e^-$ or $\\mu^+\\mu^-$. The analysis is performed on a data sample corresponding to an integrated luminosity of $711~\\mathrm{fb}^{-1}$ containing $772\\times 10^{6}$ $B\\bar B$ pairs, collected at the $\\Upsilon(4S)$ resonance with the Belle detector at the asymmetric-energy $e^+e^-$ collider KEKB. Four angular observables, $P_{4,5,6,8}'$ are extracted in five bins of the invariant mass squared of the lepton system, $q^2$. We compare our results for $P_{4,5,6,8}'$ with Standard Model predictions including the $q^2$ region in which the LHCb collaboration reported the so-called $P_5'$ anomaly.

  9. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  10. Genetic exchange versus genetic differentiation in a medium-sized inversion of Drosophila: the A2/Ast arrangements of Drosophila subobscura.

    Science.gov (United States)

    Nóbrega, Clévio; Khadem, Mahnaz; Aguadé, Montserrat; Segarra, Carmen

    2008-08-01

    Chromosomal inversion polymorphism affects nucleotide variation at loci associated with inversions. In Drosophila subobscura, a species with a rich chromosomal inversion polymorphism and the largest recombinational map so far reported in the Drosophila genus, extensive genetic structure of nucleotide variation was detected in the segment affected by the O(3) inversion, a moderately sized inversion at Muller's element E. Indeed, a strong genetic differentiation all over O(3) and no evidence of a higher genetic exchange in the center of the inversion than at breakpoints were detected. In order to ascertain, whether other polymorphic and differently sized inversions of D. subobscura also exhibited a strong genetic structure, nucleotide variation in 5 gene regions (P236, P275, P150, Sxl, and P125) located along the A(2) inversion was analyzed in A(st) and A(2) chromosomes of D. subobscura. A(2) is a medium-sized inversion at Muller's element A and forms a single inversion loop in heterokaryotypes. The lower level of variation in A(2) relative to A(st) and the significant excess of low-frequency variants at polymorphic sites indicate that nucleotide variation at A(2) is not at mutation-drift equilibrium. The closest region to an inversion breakpoint, P236, exhibits the highest level of genetic differentiation (F(ST)) and of linkage disequilibrium (LD) between arrangements and variants at nucleotide polymorphic sites. The remaining 4 regions show a higher level of genetic exchange between A(2) and A(st) chromosomes than P236, as revealed by F(ST) and LD estimates. However, significant genetic differentiation between the A(st) and A(2) arrangements was detected not only at P236 but also in the other 4 regions separated from the nearest breakpoint by 1.2-2.9 Mb. Therefore, the extent of genetic exchange between arrangements has not been high enough to homogenize nucleotide variation in the center of the A(2) inversion. A(2) can be considered a typical successful inversion

  11. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F; Leira Rey, G

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  12. IEA-R1 reactor - Spent fuel management

    International Nuclear Information System (INIS)

    Mattos, J.R.L. De

    1996-01-01

    Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

  13. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  14. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  15. Modernization and Refurbishment of the RECH-1 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

    2014-08-15

    The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

  16. Extensive utilisation of VR-1 reactor for nuclear education and training

    International Nuclear Information System (INIS)

    Rataj, J.

    2010-01-01

    The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

  17. Common reference intervals for aspartate aminotransferase (AST), alanine aminotransferase (ALT) and γ-glutamyl transferase (GGT) in serum: results from an IFCC multicenter study.

    Science.gov (United States)

    Ceriotti, Ferruccio; Henny, Joseph; Queraltó, Josep; Ziyu, Shen; Özarda, Yeşim; Chen, Baorong; Boyd, James C; Panteghini, Mauro

    2010-11-01

    Aspartate aminotransferase (AST), alanine aminotransferase (ALT) and γ-glutamyl transferase (GGT) measurements are important for the assessment of liver damage. The aim of this study was to define the reference intervals (RIs) for these enzymes in adults, paying attention to standardization of the methods used and careful selection of the reference population. AST, ALT and GGT were measured with commercial analytical systems standardized to the IFCC-recommended reference measurement systems. Three centers (two in Italy and one in China) measured their own freshly collected samples; one of these centers also measured frozen samples from the Nordic Countries RI Project and from a Turkish center. RIs were generated using non-parametric techniques from the results of 765 individuals (411 females and 354 males, 18-85 years old) selected on the basis of the results of other laboratory tests and a specific questionnaire. AST results from the four regions (Milan, Beijing, Bursa and Nordic Countries) were statistically different, but these differences were too small to be clinically relevant. Likewise, differences between the upper reference limits for genders was only 1.7 U/L (0.03 μkat/L), allowing a single RI of 11-34 U/L (0.18-0.57 μkat/L) to be defined. Interregional differences were not statistically significant for ALT, but partitioning was required due to significant gender differences. RIs for ALT were 8-41 U/L (0.13-0.68 μkat/L) for females and 9-59 U/L (0.15-0.99 μkat/L) for males, respectively. The upper reference limits for GGT from the Nordic Country population were higher than those from the other three regions and results from this group were excluded from final calculations. The GGT RIs were 6-40 U/L (0.11-0.66 μkat/L) for females and 12-68 U/L (0.20- 1.13 μkat/L) for males, respectively. For AST and ALT, the implementation of common RIs appears to be possible, because no differences between regions were observed. However, a common RI for GGT that is

  18. Antigen spot test (AST): a highly sensitive assay for the detection of antibodies

    Energy Technology Data Exchange (ETDEWEB)

    Herbrink, P; van Bussel, F J; Warnaar, S O [Rijksuniversiteit Leiden (Netherlands)

    1982-02-12

    A method is described for detection of antibodies by means of nitrocellulose or diazobenzyloxymethyl (DBM) paper on which various antigens have been spotted. The sensitivity of this antigen spot test (AST) is comparable with that of RIA and ELISA. The method requires only nanogram amounts of antigen. Since a variety of antigens can be spotted on a single piece of nitrocellulose or DBM paper, this antigen spot test is especially useful for specificity controls on antibodies.

  19. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  20. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  1. Use of the VR-1 ''Vrabec'' training reactor

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

    1994-01-01

    An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

  2. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  3. Modernization of control instrumentation and security of reactor IAN - R1

    International Nuclear Information System (INIS)

    Gonzalez, J. M.

    1993-01-01

    The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

  4. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

    The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

  5. Observation and measurement of Higgs boson decays to $WW^\\ast$ with the ATLAS detector

    CERN Document Server

    Aad, Georges; Abdallah, Jalal; Abdel Khalek, Samah; Abdinov, Ovsat; Aben, Rosemarie; Abi, Babak; Abolins, Maris; AbouZeid, Ossama; Abramowicz, Halina; Abreu, Henso; Abreu, Ricardo; Abulaiti, Yiming; Acharya, Bobby Samir; Adamczyk, Leszek; Adams, David; Adelman, Jahred; Adomeit, Stefanie; Adye, Tim; Agatonovic-Jovin, Tatjana; Aguilar-Saavedra, Juan Antonio; Agustoni, Marco; Ahlen, Steven; Ahmadov, Faig; Aielli, Giulio; Akerstedt, Henrik; Åkesson, Torsten Paul Ake; Akimoto, Ginga; Akimov, Andrei; Alberghi, Gian Luigi; Albert, Justin; Albrand, Solveig; Alconada Verzini, Maria Josefina; Aleksa, Martin; Aleksandrov, Igor; Alexa, Calin; Alexander, Gideon; Alexandre, Gauthier; Alexopoulos, Theodoros; Alhroob, Muhammad; Alimonti, Gianluca; Alio, Lion; Alison, John; Allbrooke, Benedict; Allison, Lee John; Allport, Phillip; Aloisio, Alberto; Alonso, Alejandro; Alonso, Francisco; Alpigiani, Cristiano; Altheimer, Andrew David; Alvarez Gonzalez, Barbara; Alviggi, Mariagrazia; Amako, Katsuya; Amaral Coutinho, Yara; Amelung, Christoph; Amidei, Dante; Amor Dos Santos, Susana Patricia; Amorim, Antonio; Amoroso, Simone; Amram, Nir; Amundsen, Glenn; Anastopoulos, Christos; Ancu, Lucian Stefan; Andari, Nansi; Andeen, Timothy; Anders, Christoph Falk; Anders, Gabriel; Anderson, Kelby; Andreazza, Attilio; Andrei, George Victor; Anduaga, Xabier; Angelidakis, Stylianos; Angelozzi, Ivan; Anger, Philipp; Angerami, Aaron; Anghinolfi, Francis; Anisenkov, Alexey; Anjos, Nuno; Annovi, Alberto; Antonelli, Mario; Antonov, Alexey; Antos, Jaroslav; Anulli, Fabio; Aoki, Masato; Aperio Bella, Ludovica; Arabidze, Giorgi; Arai, Yasuo; Araque, Juan Pedro; Arce, Ayana; Arduh, Francisco Anuar; Arguin, Jean-Francois; Argyropoulos, Spyridon; Arik, Metin; Armbruster, Aaron James; Arnaez, Olivier; Arnal, Vanessa; Arnold, Hannah; Arratia, Miguel; Arslan, Ozan; Artamonov, Andrei; Artoni, Giacomo; Asai, Shoji; Asbah, Nedaa; Ashkenazi, Adi; Åsman, Barbro; Asquith, Lily; Assamagan, Ketevi; Astalos, Robert; Atkinson, Markus; Atlay, Naim Bora; Auerbach, Benjamin; Augsten, Kamil; Aurousseau, Mathieu; Avolio, Giuseppe; Axen, Bradley; Azuelos, Georges; Azuma, Yuya; Baak, Max; Baas, Alessandra; Bacci, Cesare; Bachacou, Henri; Bachas, Konstantinos; Backes, Moritz; Backhaus, Malte; Badescu, Elisabeta; Bagiacchi, Paolo; Bagnaia, Paolo; Bai, Yu; Bain, Travis; Baines, John; Baker, Oliver Keith; Balek, Petr; Balli, Fabrice; Banas, Elzbieta; Banerjee, Swagato; Bannoura, Arwa A E; Bansil, Hardeep Singh; Barak, Liron; Baranov, Sergei; Barberio, Elisabetta Luigia; Barberis, Dario; Barbero, Marlon; Barillari, Teresa; Barisonzi, Marcello; Barklow, Timothy; Barlow, Nick; Barnes, Sarah Louise; Barnett, Bruce; Barnett, Michael; Barnovska, Zuzana; Baroncelli, Antonio; Barone, Gaetano; Barr, Alan; Barreiro, Fernando; Barreiro Guimarães da Costa, João; Bartoldus, Rainer; Barton, Adam Edward; Bartos, Pavol; Bartsch, Valeria; Bassalat, Ahmed; Basye, Austin; Bates, Richard; Batista, Santiago Juan; Batley, Richard; Battaglia, Marco; Battistin, Michele; Bauer, Florian; Bawa, Harinder Singh; Beacham, James Baker; Beattie, Michael David; Beau, Tristan; Beauchemin, Pierre-Hugues; Beccherle, Roberto; Bechtle, Philip; Beck, Hans Peter; Becker, Anne Kathrin; Becker, Sebastian; Beckingham, Matthew; Becot, Cyril; Beddall, Andrew; Beddall, Ayda; Bedikian, Sourpouhi; Bednyakov, Vadim; Bee, Christopher; Beemster, Lars; Beermann, Thomas; Begel, Michael; Behr, Katharina; Belanger-Champagne, Camille; Bell, Paul; Bell, William; Bella, Gideon; Bellagamba, Lorenzo; Bellerive, Alain; Bellomo, Massimiliano; Belotskiy, Konstantin; Beltramello, Olga; Benary, Odette; Benchekroun, Driss; Bendtz, Katarina; Benekos, Nektarios; Benhammou, Yan; Benhar Noccioli, Eleonora; Benitez Garcia, Jorge-Armando; Benjamin, Douglas; Bensinger, James; Bentvelsen, Stan; Berge, David; Bergeaas Kuutmann, Elin; Berger, Nicolas; Berghaus, Frank; Beringer, Jürg; Bernard, Clare; Bernard, Nathan Rogers; Bernius, Catrin; Bernlochner, Florian Urs; Berry, Tracey; Berta, Peter; Bertella, Claudia; Bertoli, Gabriele; Bertolucci, Federico; Bertsche, Carolyn; Bertsche, David; Besana, Maria Ilaria; Besjes, Geert-Jan; Bessidskaia, Olga; Bessner, Martin Florian; Besson, Nathalie; Betancourt, Christopher; Bethke, Siegfried; Bevan, Adrian John; Bhimji, Wahid; Bianchi, Riccardo-Maria; Bianchini, Louis; Bianco, Michele; Biebel, Otmar; Bieniek, Stephen Paul; Bierwagen, Katharina; Biglietti, Michela; Bilbao De Mendizabal, Javier; Bilokon, Halina; Bindi, Marcello; Binet, Sebastien; Bingul, Ahmet; Bini, Cesare; Black, Curtis; Black, James; Black, Kevin; Blackburn, Daniel; Blair, Robert; Blanchard, Jean-Baptiste; Blazek, Tomas; Bloch, Ingo; Blocker, Craig; Blum, Walter; Blumenschein, Ulrike; Bobbink, Gerjan; Bobrovnikov, Victor; Bocchetta, Simona Serena; Bocci, Andrea; Bock, Christopher; Boddy, Christopher Richard; Boehler, Michael; Boek, Thorsten Tobias; Bogaerts, Joannes Andreas; Bogdanchikov, Alexander; Bogouch, Andrei; Bohm, Christian; Boisvert, Veronique; Bold, Tomasz; Boldea, Venera; Boldyrev, Alexey; Bomben, Marco; Bona, Marcella; Boonekamp, Maarten; Borisov, Anatoly; Borissov, Guennadi; Borroni, Sara; Bortfeldt, Jonathan; Bortolotto, Valerio; Bos, Kors; Boscherini, Davide; Bosman, Martine; Boterenbrood, Hendrik; Boudreau, Joseph; Bouffard, Julian; Bouhova-Thacker, Evelina Vassileva; Boumediene, Djamel Eddine; Bourdarios, Claire; Bousson, Nicolas; Boutouil, Sara; Boveia, Antonio; Boyd, James; Boyko, Igor; Bozic, Ivan; Bracinik, Juraj; Brandt, Andrew; Brandt, Gerhard; Brandt, Oleg; Bratzler, Uwe; Brau, Benjamin; Brau, James; Braun, Helmut; Brazzale, Simone Federico; Brelier, Bertrand; Brendlinger, Kurt; Brennan, Amelia Jean; Brenner, Richard; Bressler, Shikma; Bristow, Kieran; Bristow, Timothy Michael; Britton, Dave; Brochu, Frederic; Brock, Ian; Brock, Raymond; Bronner, Johanna; Brooijmans, Gustaaf; Brooks, Timothy; Brooks, William; Brosamer, Jacquelyn; Brost, Elizabeth; Brown, Jonathan; Bruckman de Renstrom, Pawel; Bruncko, Dusan; Bruneliere, Renaud; Brunet, Sylvie; Bruni, Alessia; Bruni, Graziano; Bruschi, Marco; Bryngemark, Lene; Buanes, Trygve; Buat, Quentin; Bucci, Francesca; Buchholz, Peter; Buckley, Andrew; Buda, Stelian Ioan; Budagov, Ioulian; Buehrer, Felix; Bugge, Lars; Bugge, Magnar Kopangen; Bulekov, Oleg; Bundock, Aaron Colin; Burckhart, Helfried; Burdin, Sergey; Burghgrave, Blake; Burke, Stephen; Burmeister, Ingo; Busato, Emmanuel; Büscher, Daniel; Büscher, Volker; Bussey, Peter; Buszello, Claus-Peter; Butler, Bart; Butler, John; Butt, Aatif Imtiaz; Buttar, Craig; Butterworth, Jonathan; Butti, Pierfrancesco; Buttinger, William; Buzatu, Adrian; Byszewski, Marcin; Cabrera Urbán, Susana; Caforio, Davide; Cakir, Orhan; Calafiura, Paolo; Calandri, Alessandro; Calderini, Giovanni; Calfayan, Philippe; Caloba, Luiz; Calvet, David; Calvet, Samuel; Camacho Toro, Reina; Camarda, Stefano; Cameron, David; Caminada, Lea Michaela; Caminal Armadans, Roger; Campana, Simone; Campanelli, Mario; Campoverde, Angel; Canale, Vincenzo; Canepa, Anadi; Cano Bret, Marc; Cantero, Josu; Cantrill, Robert; Cao, Tingting; Capeans Garrido, Maria Del Mar; Caprini, Irinel; Caprini, Mihai; Capua, Marcella; Caputo, Regina; Cardarelli, Roberto; Carli, Tancredi; Carlino, Gianpaolo; Carminati, Leonardo; Caron, Sascha; Carquin, Edson; Carrillo-Montoya, German D; Carter, Janet; Carvalho, João; Casadei, Diego; Casado, Maria Pilar; Casolino, Mirkoantonio; Castaneda-Miranda, Elizabeth; Castelli, Angelantonio; Castillo Gimenez, Victoria; Castro, Nuno Filipe; Catastini, Pierluigi; Catinaccio, Andrea; Catmore, James; Cattai, Ariella; Cattani, Giordano; Caudron, Julien; Cavaliere, Viviana; Cavalli, Donatella; Cavalli-Sforza, Matteo; Cavasinni, Vincenzo; Ceradini, Filippo; Cerio, Benjamin; Cerny, Karel; Santiago Cerqueira, Augusto; Cerri, Alessandro; Cerrito, Lucio; Cerutti, Fabio; Cerv, Matevz; Cervelli, Alberto; Cetin, Serkant Ali; Chafaq, Aziz; Chakraborty, Dhiman; Chalupkova, Ina; Chang, Philip; Chapleau, Bertrand; Chapman, John Derek; Charfeddine, Driss; Charlton, Dave; Chau, Chav Chhiv; Chavez Barajas, Carlos Alberto; Cheatham, Susan; Chegwidden, Andrew; Chekanov, Sergei; Chekulaev, Sergey; Chelkov, Gueorgui; Chelstowska, Magda Anna; Chen, Chunhui; Chen, Hucheng; Chen, Karen; Chen, Liming; Chen, Shenjian; Chen, Xin; Chen, Ye; Cheng, Hok Chuen; Cheng, Yangyang; Cheplakov, Alexander; Cheremushkina, Evgenia; Cherkaoui El Moursli, Rajaa; Chernyatin, Valeriy; Cheu, Elliott; Chevalier, Laurent; Chiarella, Vitaliano; Chiefari, Giovanni; Childers, John Taylor; Chilingarov, Alexandre; Chiodini, Gabriele; Chisholm, Andrew; Chislett, Rebecca Thalatta; Chitan, Adrian; Chizhov, Mihail; Chouridou, Sofia; Chow, Bonnie Kar Bo; Chromek-Burckhart, Doris; Chu, Ming-Lee; Chudoba, Jiri; Chwastowski, Janusz; Chytka, Ladislav; Ciapetti, Guido; Ciftci, Abbas Kenan; Ciftci, Rena; Cinca, Diane; Cindro, Vladimir; Ciocio, Alessandra; Citron, Zvi Hirsh; Citterio, Mauro; Ciubancan, Mihai; Clark, Allan G; Clark, Philip James; Clarke, Robert; Cleland, Bill; Clemens, Jean-Claude; Clement, Christophe; Coadou, Yann; Cobal, Marina; Coccaro, Andrea; Cochran, James H; Coffey, Laurel; Cogan, Joshua Godfrey; Cole, Brian; Cole, Stephen; Colijn, Auke-Pieter; Collot, Johann; Colombo, Tommaso; Compostella, Gabriele; Conde Muiño, Patricia; Coniavitis, Elias; Connell, Simon Henry; Connelly, Ian; Consonni, Sofia Maria; Consorti, Valerio; Constantinescu, Serban; Conta, Claudio; Conti, Geraldine; Conventi, Francesco; Cooke, Mark; Cooper, Ben; Cooper-Sarkar, Amanda; Cooper-Smith, Neil; Copic, Katherine; Cornelissen, Thijs; Corradi, Massimo; Corriveau, Francois; Corso-Radu, Alina; Cortes-Gonzalez, Arely; Cortiana, Giorgio; Costa, Giuseppe; Costa, María José; Costanzo, Davide; Côté, David; Cottin, Giovanna; Cowan, Glen; Cox, Brian; Cranmer, Kyle; Cree, Graham; Crépé-Renaudin, Sabine; Crescioli, Francesco; Cribbs, Wayne Allen; Crispin Ortuzar, Mireia; Cristinziani, Markus; Croft, Vince; Crosetti, Giovanni; Cuhadar Donszelmann, Tulay; Cummings, Jane; Curatolo, Maria; Cuthbert, Cameron; Czirr, Hendrik; Czodrowski, Patrick; D'Auria, Saverio; D'Onofrio, Monica; Da Cunha Sargedas De Sousa, Mario Jose; Da Via, Cinzia; Dabrowski, Wladyslaw; Dafinca, Alexandru; Dai, Tiesheng; Dale, Orjan; Dallaire, Frederick; Dallapiccola, Carlo; Dam, Mogens; Daniells, Andrew Christopher; Danninger, Matthias; Dano Hoffmann, Maria; Dao, Valerio; Darbo, Giovanni; Darmora, Smita; Dassoulas, James; Dattagupta, Aparajita; Davey, Will; David, Claire; Davidek, Tomas; Davies, Eleanor; Davies, Merlin; Davignon, Olivier; Davison, Adam; Davison, Peter; Davygora, Yuriy; Dawe, Edmund; Dawson, Ian; Daya-Ishmukhametova, Rozmin; De, Kaushik; de Asmundis, Riccardo; De Castro, Stefano; De Cecco, Sandro; De Groot, Nicolo; de Jong, Paul; De la Torre, Hector; De Lorenzi, Francesco; De Nooij, Lucie; De Pedis, Daniele; De Salvo, Alessandro; De Sanctis, Umberto; De Santo, Antonella; De Vivie De Regie, Jean-Baptiste; Dearnaley, William James; Debbe, Ramiro; Debenedetti, Chiara; Dechenaux, Benjamin; Dedovich, Dmitri; Deigaard, Ingrid; Del Peso, Jose; Del Prete, Tarcisio; Deliot, Frederic; Delitzsch, Chris Malena; Deliyergiyev, Maksym; Dell'Acqua, Andrea; Dell'Asta, Lidia; Dell'Orso, Mauro; Della Pietra, Massimo; della Volpe, Domenico; Delmastro, Marco; Delsart, Pierre-Antoine; Deluca, Carolina; DeMarco, David; Demers, Sarah; Demichev, Mikhail; Demilly, Aurelien; Denisov, Sergey; Derendarz, Dominik; Derkaoui, Jamal Eddine; Derue, Frederic; Dervan, Paul; Desch, Klaus Kurt; Deterre, Cecile; Deviveiros, Pier-Olivier; Dewhurst, Alastair; Dhaliwal, Saminder; Di Ciaccio, Anna; Di Ciaccio, Lucia; Di Domenico, Antonio; Di Donato, Camilla; Di Girolamo, Alessandro; Di Girolamo, Beniamino; Di Mattia, Alessandro; Di Micco, Biagio; Di Nardo, Roberto; Di Simone, Andrea; Di Sipio, Riccardo; Di Valentino, David; Dias, Flavia; Diaz, Marco Aurelio; Diehl, Edward; Dietrich, Janet; Dietzsch, Thorsten; Diglio, Sara; Dimitrievska, Aleksandra; Dingfelder, Jochen; Dita, Petre; Dita, Sanda; Dittus, Fridolin; Djama, Fares; Djobava, Tamar; Djuvsland, Julia Isabell; Barros do Vale, Maria Aline; Dobos, Daniel; Doglioni, Caterina; Doherty, Tom; Dohmae, Takeshi; Dolejsi, Jiri; Dolezal, Zdenek; Dolgoshein, Boris; Donadelli, Marisilvia; Donati, Simone; Dondero, Paolo; Donini, Julien; Dopke, Jens; Doria, Alessandra; Dova, Maria-Teresa; Doyle, Tony; Dris, Manolis; Dubbert, Jörg; Dube, Sourabh; Dubreuil, Emmanuelle; Duchovni, Ehud; Duckeck, Guenter; Ducu, Otilia Anamaria; Duda, Dominik; Dudarev, Alexey; Dudziak, Fanny; Duflot, Laurent; Duguid, Liam; Dührssen, Michael; Dunford, Monica; Duran Yildiz, Hatice; Düren, Michael; Durglishvili, Archil; Duschinger, Dirk; Dwuznik, Michal; Dyndal, Mateusz; Edson, William; Edwards, Nicholas Charles; Ehrenfeld, Wolfgang; Eifert, Till; Eigen, Gerald; Einsweiler, Kevin; Ekelof, Tord; El Kacimi, Mohamed; Ellert, Mattias; Elles, Sabine; Ellinghaus, Frank; Elliot, Alison; Ellis, Nicolas; Elmsheuser, Johannes; Elsing, Markus; Emeliyanov, Dmitry; Enari, Yuji; Endner, Oliver Chris; Endo, Masaki; Engelmann, Roderich; Erdmann, Johannes; Ereditato, Antonio; Eriksson, Daniel; Ernis, Gunar; Ernst, Jesse; Ernst, Michael; Ernwein, Jean; Errede, Steven; Ertel, Eugen; Escalier, Marc; Esch, Hendrik; Escobar, Carlos; Esposito, Bellisario; Etienvre, Anne-Isabelle; Etzion, Erez; Evans, Hal; Ezhilov, Alexey; Fabbri, Laura; Facini, Gabriel; Fakhrutdinov, Rinat; Falciano, Speranza; Falla, Rebecca Jane; Faltova, Jana; Fang, Yaquan; Fanti, Marcello; Farbin, Amir; Farilla, Addolorata; Farooque, Trisha; Farrell, Steven; Farrington, Sinead; Farthouat, Philippe; Fassi, Farida; Fassnacht, Patrick; Fassouliotis, Dimitrios; Favareto, Andrea; Fayard, Louis; Federic, Pavol; Fedin, Oleg; Fedorko, Wojciech; Feigl, Simon; Feligioni, Lorenzo; Feng, Cunfeng; Feng, Eric; Feng, Haolu; Fenyuk, Alexander; Fernandez Martinez, Patricia; Fernandez Perez, Sonia; Ferrag, Samir; Ferrando, James; Ferrari, Arnaud; Ferrari, Pamela; Ferrari, Roberto; Ferreira de Lima, Danilo Enoque; Ferrer, Antonio; Ferrere, Didier; Ferretti, Claudio; Ferretto Parodi, Andrea; Fiascaris, Maria; Fiedler, Frank; Filipčič, Andrej; Filipuzzi, Marco; Filthaut, Frank; Fincke-Keeler, Margret; Finelli, Kevin Daniel; Fiolhais, Miguel; Fiorini, Luca; Firan, Ana; Fischer, Adam; Fischer, Julia; Fisher, Wade Cameron; Fitzgerald, Eric Andrew; Flechl, Martin; Fleck, Ivor; Fleischmann, Philipp; Fleischmann, Sebastian; Fletcher, Gareth Thomas; Fletcher, Gregory; Flick, Tobias; Floderus, Anders; Flores Castillo, Luis; Flowerdew, Michael; Formica, Andrea; Forti, Alessandra; Fournier, Daniel; Fox, Harald; Fracchia, Silvia; Francavilla, Paolo; Franchini, Matteo; Franchino, Silvia; Francis, David; Franconi, Laura; Franklin, Melissa; Fraternali, Marco; French, Sky; Friedrich, Conrad; Friedrich, Felix; Froidevaux, Daniel; Frost, James; Fukunaga, Chikara; Fullana Torregrosa, Esteban; Fulsom, Bryan Gregory; Fuster, Juan; Gabaldon, Carolina; Gabizon, Ofir; Gabrielli, Alessandro; Gabrielli, Andrea; Gadatsch, Stefan; Gadomski, Szymon; Gagliardi, Guido; Gagnon, Pauline; Galea, Cristina; Galhardo, Bruno; Gallas, Elizabeth; Gallop, Bruce; Gallus, Petr; Galster, Gorm Aske Gram Krohn; Gan, KK; Gao, Jun; Gao, Yongsheng; Garay Walls, Francisca; Garberson, Ford; García, Carmen; García Navarro, José Enrique; Garcia-Sciveres, Maurice; Gardner, Robert; Garelli, Nicoletta; Garonne, Vincent; Gatti, Claudio; Gaudio, Gabriella; Gaur, Bakul; Gauthier, Lea; Gauzzi, Paolo; Gavrilenko, Igor; Gay, Colin; Gaycken, Goetz; Gazis, Evangelos; Ge, Peng; Gecse, Zoltan; Gee, Norman; Geerts, Daniël Alphonsus Adrianus; Geich-Gimbel, Christoph; Gellerstedt, Karl; Gemme, Claudia; Gemmell, Alistair; Genest, Marie-Hélène; Gentile, Simonetta; George, Matthias; George, Simon; Gerbaudo, Davide; Gershon, Avi; Ghazlane, Hamid; Ghodbane, Nabil; Giacobbe, Benedetto; Giagu, Stefano; Giangiobbe, Vincent; Giannetti, Paola; Gianotti, Fabiola; Gibbard, Bruce; Gibson, Stephen; Gilchriese, Murdock; Gillam, Thomas; Gillberg, Dag; Gilles, Geoffrey; Gingrich, Douglas; Giokaris, Nikos; Giordani, MarioPaolo; Giordano, Raffaele; Giorgi, Filippo Maria; Giorgi, Francesco Michelangelo; Giraud, Pierre-Francois; Giugni, Danilo; Giuliani, Claudia; Giulini, Maddalena; Gjelsten, Børge Kile; Gkaitatzis, Stamatios; Gkialas, Ioannis; Gkougkousis, Evangelos Leonidas; Gladilin, Leonid; Glasman, Claudia; Glatzer, Julian; Glaysher, Paul; Glazov, Alexandre; Glonti, George; Goblirsch-Kolb, Maximilian; Goddard, Jack Robert; Godlewski, Jan; Goldfarb, Steven; Golling, Tobias; Golubkov, Dmitry; Gomes, Agostinho; Gomez Fajardo, Luz Stella; Gonçalo, Ricardo; Goncalves Pinto Firmino Da Costa, Joao; Gonella, Laura; González de la Hoz, Santiago; Gonzalez Parra, Garoe; Gonzalez-Sevilla, Sergio; Goossens, Luc; Gorbounov, Petr Andreevich; Gordon, Howard; Gorelov, Igor; Gorini, Benedetto; Gorini, Edoardo; Gorišek, Andrej; Gornicki, Edward; Goshaw, Alfred; Gössling, Claus; Gostkin, Mikhail Ivanovitch; Gouighri, Mohamed; Goujdami, Driss; Goulette, Marc Phillippe; Goussiou, Anna; Goy, Corinne; Grabas, Herve Marie Xavier; Graber, Lars; Grabowska-Bold, Iwona; Grafström, Per; Grahn, Karl-Johan; Gramling, Johanna; Gramstad, Eirik; Grancagnolo, Sergio; Grassi, Valerio; Gratchev, Vadim; Gray, Heather; Graziani, Enrico; Grebenyuk, Oleg; Greenwood, Zeno Dixon; Gregersen, Kristian; Gregor, Ingrid-Maria; Grenier, Philippe; Griffiths, Justin; Grillo, Alexander; Grimm, Kathryn; Grinstein, Sebastian; Gris, Philippe Luc Yves; Grishkevich, Yaroslav; Grivaz, Jean-Francois; Grohs, Johannes Philipp; Grohsjean, Alexander; Gross, Eilam; Grosse-Knetter, Joern; Grossi, Giulio Cornelio; Grout, Zara Jane; Guan, Liang; Guenther, Jaroslav; Guescini, Francesco; Guest, Daniel; Gueta, Orel; Guicheney, Christophe; Guido, Elisa; Guillemin, Thibault; Guindon, Stefan; Gul, Umar; Gumpert, Christian; Guo, Jun; Gupta, Shaun; Gutierrez, Phillip; Gutierrez Ortiz, Nicolas Gilberto; Gutschow, Christian; Guttman, Nir; Guyot, Claude; Gwenlan, Claire; Gwilliam, Carl; Haas, Andy; Haber, Carl; Hadavand, Haleh Khani; Haddad, Nacim; Haefner, Petra; Hageböck, Stephan; Hajduk, Zbigniew; Hakobyan, Hrachya; Haleem, Mahsana; Haley, Joseph; Hall, David; Halladjian, Garabed; Hallewell, Gregory David; Hamacher, Klaus; Hamal, Petr; Hamano, Kenji; Hamer, Matthias; Hamilton, Andrew; Hamilton, Samuel; Hamity, Guillermo Nicolas; Hamnett, Phillip George; Han, Liang; Hanagaki, Kazunori; Hanawa, Keita; Hance, Michael; Hanke, Paul; Hanna, Remie; Hansen, Jørgen Beck; Hansen, Jorn Dines; Hansen, Peter Henrik; Hara, Kazuhiko; Hard, Andrew; Harenberg, Torsten; Hariri, Faten; Harkusha, Siarhei; Harrington, Robert; Harrison, Paul Fraser; Hartjes, Fred; Hasegawa, Makoto; Hasegawa, Satoshi; Hasegawa, Yoji; Hasib, A; Hassani, Samira; Haug, Sigve; Hauschild, Michael; Hauser, Reiner; Havranek, Miroslav; Hawkes, Christopher; Hawkings, Richard John; Hawkins, Anthony David; Hayashi, Takayasu; Hayden, Daniel; Hays, Chris; Hays, Jonathan Michael; Hayward, Helen; Haywood, Stephen; Head, Simon; Heck, Tobias; Hedberg, Vincent; Heelan, Louise; Heim, Sarah; Heim, Timon; Heinemann, Beate; Heinrich, Lukas; Hejbal, Jiri; Helary, Louis; Heller, Matthieu; Hellman, Sten; Hellmich, Dennis; Helsens, Clement; Henderson, James; Henderson, Robert; Heng, Yang; Hengler, Christopher; Henrichs, Anna; Henriques Correia, Ana Maria; Henrot-Versille, Sophie; Herbert, Geoffrey Henry; Hernández Jiménez, Yesenia; Herrberg-Schubert, Ruth; Herten, Gregor; Hertenberger, Ralf; Hervas, Luis; Hesketh, Gavin Grant; Hessey, Nigel; Hickling, Robert; Higón-Rodriguez, Emilio; Hill, Ewan; Hill, John; Hiller, Karl Heinz; Hillier, Stephen; Hinchliffe, Ian; Hines, Elizabeth; Hinman, Rachel Reisner; Hirose, Minoru; Hirschbuehl, Dominic; Hobbs, John; Hod, Noam; Hodgkinson, Mark; Hodgson, Paul; Hoecker, Andreas; Hoeferkamp, Martin; Hoenig, Friedrich; Hoffmann, Dirk; Hohlfeld, Marc; Holmes, Tova Ray; Hong, Tae Min; Hooft van Huysduynen, Loek; Hopkins, Walter; Horii, Yasuyuki; Horton, Arthur James; Hostachy, Jean-Yves; Hou, Suen; Hoummada, Abdeslam; Howard, Jacob; Howarth, James; Hrabovsky, Miroslav; Hristova, Ivana; Hrivnac, Julius; Hryn'ova, Tetiana; Hrynevich, Aliaksei; Hsu, Catherine; Hsu, Pai-hsien Jennifer; Hsu, Shih-Chieh; Hu, Diedi; Hu, Xueye; Huang, Yanping; Hubacek, Zdenek; Hubaut, Fabrice; Huegging, Fabian; Huffman, Todd Brian; Hughes, Emlyn; Hughes, Gareth; Huhtinen, Mika; Hülsing, Tobias Alexander; Hurwitz, Martina; Huseynov, Nazim; Huston, Joey; Huth, John; Iacobucci, Giuseppe; Iakovidis, Georgios; Ibragimov, Iskander; Iconomidou-Fayard, Lydia; Ideal, Emma; Idrissi, Zineb; Iengo, Paolo; Igonkina, Olga; Iizawa, Tomoya; Ikegami, Yoichi; Ikematsu, Katsumasa; Ikeno, Masahiro; Ilchenko, Iurii; Iliadis, Dimitrios; Ilic, Nikolina; Inamaru, Yuki; Ince, Tayfun; Ioannou, Pavlos; Iodice, Mauro; Iordanidou, Kalliopi; Ippolito, Valerio; Irles Quiles, Adrian; Isaksson, Charlie; Ishino, Masaya; Ishitsuka, Masaki; Ishmukhametov, Renat; Issever, Cigdem; Istin, Serhat; Iturbe Ponce, Julia Mariana; Iuppa, Roberto; Ivarsson, Jenny; Iwanski, Wieslaw; Iwasaki, Hiroyuki; Izen, Joseph; Izzo, Vincenzo; Jackson, Brett; Jackson, Matthew; Jackson, Paul; Jaekel, Martin; Jain, Vivek; Jakobs, Karl; Jakobsen, Sune; Jakoubek, Tomas; Jakubek, Jan; Jamin, David Olivier; Jana, Dilip; Jansen, Eric; Janssen, Jens; Janus, Michel; Jarlskog, Göran; Javadov, Namig; Javůrek, Tomáš; Jeanty, Laura; Jejelava, Juansher; Jeng, Geng-yuan; Jennens, David; Jenni, Peter; Jentzsch, Jennifer; Jeske, Carl; Jézéquel, Stéphane; Ji, Haoshuang; Jia, Jiangyong; Jiang, Yi; Jimenez Belenguer, Marcos; Jin, Shan; Jinaru, Adam; Jinnouchi, Osamu; Joergensen, Morten Dam; Johansson, Per; Johns, Kenneth; Jon-And, Kerstin; Jones, Graham; Jones, Roger; Jones, Tim; Jongmanns, Jan; Jorge, Pedro; Joshi, Kiran Daniel; Jovicevic, Jelena; Ju, Xiangyang; Jung, Christian; Jussel, Patrick; Juste Rozas, Aurelio; Kaci, Mohammed; Kaczmarska, Anna; Kado, Marumi; Kagan, Harris; Kagan, Michael; Kajomovitz, Enrique; Kalderon, Charles William; Kama, Sami; Kamenshchikov, Andrey; Kanaya, Naoko; Kaneda, Michiru; Kaneti, Steven; Kantserov, Vadim; Kanzaki, Junichi; Kaplan, Benjamin; Kapliy, Anton; Kar, Deepak; Karakostas, Konstantinos; Karamaoun, Andrew; Karastathis, Nikolaos; Kareem, Mohammad Jawad; Karnevskiy, Mikhail; Karpov, Sergey; Karpova, Zoya; Karthik, Krishnaiyengar; Kartvelishvili, Vakhtang; Karyukhin, Andrey; Kashif, Lashkar; Kasieczka, Gregor; Kass, Richard; Kastanas, Alex; Kataoka, Yousuke; Katre, Akshay; Katzy, Judith; Kaushik, Venkatesh; Kawagoe, Kiyotomo; Kawamoto, Tatsuo; Kawamura, Gen; Kazama, Shingo; Kazanin, Vassili; Kazarinov, Makhail; Keeler, Richard; Kehoe, Robert; Keil, Markus; Keller, John; Kempster, Jacob Julian; Keoshkerian, Houry; Kepka, Oldrich; Kerševan, Borut Paul; Kersten, Susanne; Kessoku, Kohei; Keung, Justin; Keyes, Robert; Khalil-zada, Farkhad; Khandanyan, Hovhannes; Khanov, Alexander; Kharlamov, Alexey; Khodinov, Alexander; Khomich, Andrei; Khoo, Teng Jian; Khoriauli, Gia; Khovanskiy, Valery; Khramov, Evgeniy; Khubua, Jemal; Kim, Hee Yeun; Kim, Hyeon Jin; Kim, Shinhong; Kimura, Naoki; Kind, Oliver; King, Barry; King, Matthew; King, Robert Steven Beaufoy; King, Samuel Burton; Kirk, Julie; Kiryunin, Andrey; Kishimoto, Tomoe; Kisielewska, Danuta; Kiss, Florian; Kiuchi, Kenji; Kladiva, Eduard; Klein, Max; Klein, Uta; Kleinknecht, Konrad; Klimek, Pawel; Klimentov, Alexei; Klingenberg, Reiner; Klinger, Joel Alexander; Klioutchnikova, Tatiana; Klok, Peter; Kluge, Eike-Erik; Kluit, Peter; Kluth, Stefan; Kneringer, Emmerich; Knoops, Edith; Knue, Andrea; Kobayashi, Dai; Kobayashi, Tomio; Kobel, Michael; Kocian, Martin; Kodys, Peter; Koffas, Thomas; Koffeman, Els; Kogan, Lucy Anne; Kohlmann, Simon; Kohout, Zdenek; Kohriki, Takashi; Koi, Tatsumi; Kolanoski, Hermann; Koletsou, Iro; Koll, James; Komar, Aston; Komori, Yuto; Kondo, Takahiko; Kondrashova, Nataliia; Köneke, Karsten; König, Adriaan; König, Sebastian; Kono, Takanori; Konoplich, Rostislav; Konstantinidis, Nikolaos; Kopeliansky, Revital; Koperny, Stefan; Köpke, Lutz; Kopp, Anna Katharina; Korcyl, Krzysztof; Kordas, Kostantinos; Korn, Andreas; Korol, Aleksandr; Korolkov, Ilya; Korolkova, Elena; Korotkov, Vladislav; Kortner, Oliver; Kortner, Sandra; Kostyukhin, Vadim; Kotov, Vladislav; Kotwal, Ashutosh; Kourkoumeli-Charalampidi, Athina; Kourkoumelis, Christine; Kouskoura, Vasiliki; Koutsman, Alex; Kowalewski, Robert Victor; Kowalski, Tadeusz; Kozanecki, Witold; Kozhin, Anatoly; Kramarenko, Viktor; Kramberger, Gregor; Krasnopevtsev, Dimitriy; Krasznahorkay, Attila; Kraus, Jana; Kravchenko, Anton; Kreiss, Sven; Kretz, Moritz; Kretzschmar, Jan; Kreutzfeldt, Kristof; Krieger, Peter; Krizka, Karol; Kroeninger, Kevin; Kroha, Hubert; Kroll, Joe; Kroseberg, Juergen; Krstic, Jelena; Kruchonak, Uladzimir; Krüger, Hans; Krumnack, Nils; Krumshteyn, Zinovii; Kruse, Amanda; Kruse, Mark; Kruskal, Michael; Kubota, Takashi; Kucuk, Hilal; Kuday, Sinan; Kuehn, Susanne; Kugel, Andreas; Kuger, Fabian; Kuhl, Andrew; Kuhl, Thorsten; Kukhtin, Victor; Kulchitsky, Yuri; Kuleshov, Sergey; Kuna, Marine; Kunigo, Takuto; Kupco, Alexander; Kurashige, Hisaya; Kurochkin, Yurii; Kurumida, Rie; Kus, Vlastimil; Kuwertz, Emma Sian; Kuze, Masahiro; Kvita, Jiri; Kyriazopoulos, Dimitrios; La Rosa, Alessandro; La Rotonda, Laura; Lacasta, Carlos; Lacava, Francesco; Lacey, James; Lacker, Heiko; Lacour, Didier; Lacuesta, Vicente Ramón; Ladygin, Evgueni; Lafaye, Remi; Laforge, Bertrand; Lagouri, Theodota; Lai, Stanley; Laier, Heiko; Lambourne, Luke; Lammers, Sabine; Lampen, Caleb; Lampl, Walter; Lançon, Eric; Landgraf, Ulrich; Landon, Murrough; Lang, Valerie Susanne; Lankford, Andrew; Lanni, Francesco; Lantzsch, Kerstin; Laplace, Sandrine; Lapoire, Cecile; Laporte, Jean-Francois; Lari, Tommaso; Lasagni Manghi, Federico; Lassnig, Mario; Laurelli, Paolo; Lavrijsen, Wim; Law, Alexander; Laycock, Paul; Le Dortz, Olivier; Le Guirriec, Emmanuel; Le Menedeu, Eve; LeCompte, Thomas; Ledroit-Guillon, Fabienne Agnes Marie; Lee, Claire Alexandra; Lee, Hurng-Chun; Lee, Shih-Chang; Lee, Lawrence; Lefebvre, Guillaume; Lefebvre, Michel; Legger, Federica; Leggett, Charles; Lehan, Allan; Lehmann Miotto, Giovanna; Lei, Xiaowen; Leight, William Axel; Leisos, Antonios; Leister, Andrew Gerard; Leite, Marco Aurelio Lisboa; Leitner, Rupert; Lellouch, Daniel; Lemmer, Boris; Leney, Katharine; Lenz, Tatjana; Lenzen, Georg; Lenzi, Bruno; Leone, Robert; Leone, Sandra; Leonidopoulos, Christos; Leontsinis, Stefanos; Leroy, Claude; Lester, Christopher; Lester, Christopher Michael; Levchenko, Mikhail; Levêque, Jessica; Levin, Daniel; Levinson, Lorne; Levy, Mark; Lewis, Adrian; Leyko, Agnieszka; Leyton, Michael; Li, Bing; Li, Bo; Li, Haifeng; Li, Ho Ling; Li, Lei; Li, Liang; Li, Shu; Li, Yichen; Liang, Zhijun; Liao, Hongbo; Liberti, Barbara; Lichard, Peter; Lie, Ki; Liebal, Jessica; Liebig, Wolfgang; Limbach, Christian; Limosani, Antonio; Lin, Simon; Lin, Tai-Hua; Linde, Frank; Lindquist, Brian Edward; Linnemann, James; Lipeles, Elliot; Lipniacka, Anna; Lisovyi, Mykhailo; Liss, Tony; Lissauer, David; Lister, Alison; Litke, Alan; Liu, Bo; Liu, Dong; Liu, Jian; Liu, Jianbei; Liu, Kun; Liu, Lulu; Liu, Miaoyuan; Liu, Minghui; Liu, Yanwen; Livan, Michele; Lleres, Annick; Llorente Merino, Javier; Lloyd, Stephen; Lo Sterzo, Francesco; Lobodzinska, Ewelina; Loch, Peter; Lockman, William; Loebinger, Fred; Loevschall-Jensen, Ask Emil; Loginov, Andrey; Lohse, Thomas; Lohwasser, Kristin; Lokajicek, Milos; Long, Brian Alexander; Long, Jonathan; Long, Robin Eamonn; Looper, Kristina Anne; Lopes, Lourenco; Lopez Mateos, David; Lopez Paredes, Brais; Lopez Paz, Ivan; Lorenz, Jeanette; Lorenzo Martinez, Narei; Losada, Marta; Loscutoff, Peter; Lou, XinChou; Lounis, Abdenour; Love, Jeremy; Love, Peter; Lowe, Andrew; Lu, Feng; Lu, Nan; Lubatti, Henry; Luci, Claudio; Lucotte, Arnaud; Luehring, Frederick; Lukas, Wolfgang; Luminari, Lamberto; Lundberg, Olof; Lund-Jensen, Bengt; Lungwitz, Matthias; Lynn, David; Lysak, Roman; Lytken, Else; Ma, Hong; Ma, Lian Liang; Maccarrone, Giovanni; Macchiolo, Anna; Machado Miguens, Joana; Macina, Daniela; Madaffari, Daniele; Madar, Romain; Maddocks, Harvey Jonathan; Mader, Wolfgang; Madsen, Alexander; Maeno, Mayuko; Maeno, Tadashi; Maevskiy, Artem; Magradze, Erekle; Mahboubi, Kambiz; Mahlstedt, Joern; Mahmoud, Sara; Maiani, Camilla; Maidantchik, Carmen; Maier, Andreas Alexander; Maio, Amélia; Majewski, Stephanie; Makida, Yasuhiro; Makovec, Nikola; Mal, Prolay; Malaescu, Bogdan; Malecki, Pawel; Maleev, Victor; Malek, Fairouz; Mallik, Usha; Malon, David; Malone, Caitlin; Maltezos, Stavros; Malyshev, Vladimir; Malyukov, Sergei; Mamuzic, Judita; Mandelli, Beatrice; Mandelli, Luciano; Mandić, Igor; Mandrysch, Rocco; Maneira, José; Manfredini, Alessandro; Manhaes de Andrade Filho, Luciano; Manjarres Ramos, Joany; Mann, Alexander; Manning, Peter; Manousakis-Katsikakis, Arkadios; Mansoulie, Bruno; Mantifel, Rodger; Mantoani, Matteo; Mapelli, Livio; March, Luis; Marchand, Jean-Francois; Marchiori, Giovanni; Marcisovsky, Michal; Marino, Christopher; Marjanovic, Marija; Marroquim, Fernando; Marsden, Stephen Philip; Marshall, Zach; Marti, Lukas Fritz; Marti-Garcia, Salvador; Martin, Brian; Martin, Brian Thomas; Martin, Tim; Martin, Victoria Jane; Martin dit Latour, Bertrand; Martinez, Homero; Martinez, Mario; Martin-Haugh, Stewart; Martyniuk, Alex; Marx, Marilyn; Marzano, Francesco; Marzin, Antoine; Masetti, Lucia; Mashimo, Tetsuro; Mashinistov, Ruslan; Masik, Jiri; Maslennikov, Alexey; Massa, Ignazio; Massa, Lorenzo; Massol, Nicolas; Mastrandrea, Paolo; Mastroberardino, Anna; Masubuchi, Tatsuya; Mättig, Peter; Mattmann, Johannes; Maurer, Julien; Maxfield, Stephen; Maximov, Dmitriy; Mazini, Rachid; Mazza, Simone Michele; Mazzaferro, Luca; Mc Goldrick, Garrin; Mc Kee, Shawn Patrick; McCarn, Allison; McCarthy, Robert; McCarthy, Tom; McCubbin, Norman; McFarlane, Kenneth; Mcfayden, Josh; Mchedlidze, Gvantsa; McMahon, Steve; McPherson, Robert; Mechnich, Joerg; Medinnis, Michael; Meehan, Samuel; Mehlhase, Sascha; Mehta, Andrew; Meier, Karlheinz; Meineck, Christian; Meirose, Bernhard; Melachrinos, Constantinos; Mellado Garcia, Bruce Rafael; Meloni, Federico; Mengarelli, Alberto; Menke, Sven; Meoni, Evelin; Mercurio, Kevin Michael; Mergelmeyer, Sebastian; Meric, Nicolas; Mermod, Philippe; Merola, Leonardo; Meroni, Chiara; Merritt, Frank; Merritt, Hayes; Messina, Andrea; Metcalfe, Jessica; Mete, Alaettin Serhan; Meyer, Carsten; Meyer, Christopher; Meyer, Jean-Pierre; Meyer, Jochen; Middleton, Robin; Migas, Sylwia; Miglioranzi, Silvia; Mijović, Liza; Mikenberg, Giora; Mikestikova, Marcela; Mikuž, Marko; Milic, Adriana; Miller, David; Mills, Corrinne; Milov, Alexander; Milstead, David; Minaenko, Andrey; Minami, Yuto; Minashvili, Irakli; Mincer, Allen; Mindur, Bartosz; Mineev, Mikhail; Ming, Yao; Mir, Lluisa-Maria; Mirabelli, Giovanni; Mitani, Takashi; Mitrevski, Jovan; Mitsou, Vasiliki A; Miucci, Antonio; Miyagawa, Paul; Mjörnmark, Jan-Ulf; Moa, Torbjoern; Mochizuki, Kazuya; Mohapatra, Soumya; Mohr, Wolfgang; Molander, Simon; Moles-Valls, Regina; Mönig, Klaus; Monini, Caterina; Monk, James; Monnier, Emmanuel; Montejo Berlingen, Javier; Monticelli, Fernando; Monzani, Simone; Moore, Roger; Morange, Nicolas; Moreno, Deywis; Moreno Llácer, María; Morettini, Paolo; Morgenstern, Marcus; Mori, Daniel; Morii, Masahiro; Morisbak, Vanja; Moritz, Sebastian; Morley, Anthony Keith; Mornacchi, Giuseppe; Morris, John; Morton, Alexander; Morvaj, Ljiljana; Moser, Hans-Guenther; Mosidze, Maia; Moss, Josh; Motohashi, Kazuki; Mount, Richard; Mountricha, Eleni; Mouraviev, Sergei; Moyse, Edward; Muanza, Steve; Mudd, Richard; Mueller, Felix; Mueller, James; Mueller, Klemens; Mueller, Thibaut; Muenstermann, Daniel; Mullen, Paul; Munwes, Yonathan; Murillo Quijada, Javier Alberto; Murray, Bill; Musheghyan, Haykuhi; Musto, Elisa; Myagkov, Alexey; Myska, Miroslav; Nackenhorst, Olaf; Nadal, Jordi; Nagai, Koichi; Nagai, Ryo; Nagai, Yoshikazu; Nagano, Kunihiro; Nagarkar, Advait; Nagasaka, Yasushi; Nagata, Kazuki; Nagel, Martin; Nairz, Armin Michael; Nakahama, Yu; Nakamura, Koji; Nakamura, Tomoaki; Nakano, Itsuo; Namasivayam, Harisankar; Nanava, Gizo; Naranjo Garcia, Roger Felipe; Narayan, Rohin; Nattermann, Till; Naumann, Thomas; Navarro, Gabriela; Nayyar, Ruchika; Neal, Homer; Nechaeva, Polina; Neep, Thomas James; Nef, Pascal Daniel; Negri, Andrea; Negri, Guido; Negrini, Matteo; Nektarijevic, Snezana; Nellist, Clara; Nelson, Andrew; Nelson, Timothy Knight; Nemecek, Stanislav; Nemethy, Peter; Nepomuceno, Andre Asevedo; Nessi, Marzio; Neubauer, Mark; Neumann, Manuel; Neves, Ricardo; Nevski, Pavel; Newman, Paul; Nguyen, Duong Hai; Nickerson, Richard; Nicolaidou, Rosy; Nicquevert, Bertrand; Nielsen, Jason; Nikiforou, Nikiforos; Nikiforov, Andriy; Nikolaenko, Vladimir; Nikolic-Audit, Irena; Nikolics, Katalin; Nikolopoulos, Konstantinos; Nilsson, Paul; Ninomiya, Yoichi; Nisati, Aleandro; Nisius, Richard; Nobe, Takuya; Nomachi, Masaharu; Nomidis, Ioannis; Norberg, Scarlet; Nordberg, Markus; Novgorodova, Olga; Nowak, Sebastian; Nozaki, Mitsuaki; Nozka, Libor; Ntekas, Konstantinos; Nunes Hanninger, Guilherme; Nunnemann, Thomas; Nurse, Emily; Nuti, Francesco; O'Brien, Brendan Joseph; O'grady, Fionnbarr; O'Neil, Dugan; O'Shea, Val; Oakham, Gerald; Oberlack, Horst; Obermann, Theresa; Ocariz, Jose; Ochi, Atsuhiko; Ochoa, Ines; Oda, Susumu; Odaka, Shigeru; Ogren, Harold; Oh, Alexander; Oh, Seog; Ohm, Christian; Ohman, Henrik; Oide, Hideyuki; Okamura, Wataru; Okawa, Hideki; Okumura, Yasuyuki; Okuyama, Toyonobu; Olariu, Albert; Olchevski, Alexander; Olivares Pino, Sebastian Andres; Oliveira Damazio, Denis; Oliver Garcia, Elena; Olszewski, Andrzej; Olszowska, Jolanta; Onofre, António; Onyisi, Peter; Oram, Christopher; Oreglia, Mark; Oren, Yona; Orestano, Domizia; Orlando, Nicola; Oropeza Barrera, Cristina; Orr, Robert; Osculati, Bianca; Ospanov, Rustem; Otero y Garzon, Gustavo; Otono, Hidetoshi; Ouchrif, Mohamed; Ouellette, Eric; Ould-Saada, Farid; Ouraou, Ahmimed; Oussoren, Koen Pieter; Ouyang, Qun; Ovcharova, Ana; Owen, Mark; Ozcan, Veysi Erkcan; Ozturk, Nurcan; Pachal, Katherine; Pacheco Pages, Andres; Padilla Aranda, Cristobal; Pagáčová, Martina; Pagan Griso, Simone; Paganis, Efstathios; Pahl, Christoph; Paige, Frank; Pais, Preema; Pajchel, Katarina; Palacino, Gabriel; Palestini, Sandro; Palka, Marek; Pallin, Dominique; Palma, Alberto; Palmer, Jody; Pan, Yibin; Panagiotopoulou, Evgenia; Panduro Vazquez, William; Pani, Priscilla; Panikashvili, Natalia; Panitkin, Sergey; Pantea, Dan; Paolozzi, Lorenzo; Papadopoulou, Theodora; Papageorgiou, Konstantinos; Paramonov, Alexander; Paredes Hernandez, Daniela; Parker, Michael Andrew; Parodi, Fabrizio; Parsons, John; Parzefall, Ulrich; Pasqualucci, Enrico; Passaggio, Stefano; Passeri, Antonio; Pastore, Fernanda; Pastore, Francesca; Pásztor, Gabriella; Pataraia, Sophio; Patel, Nikhul; Pater, Joleen; Patricelli, Sergio; Pauly, Thilo; Pearce, James; Pedersen, Lars Egholm; Pedersen, Maiken; Pedraza Lopez, Sebastian; Pedro, Rute; Peleganchuk, Sergey; Pelikan, Daniel; Peng, Haiping; Penning, Bjoern; Penwell, John; Perepelitsa, Dennis; Perez Codina, Estel; Pérez García-Estañ, María Teresa; Perini, Laura; Pernegger, Heinz; Perrella, Sabrina; Peschke, Richard; Peshekhonov, Vladimir; Peters, Krisztian; Peters, Yvonne; Petersen, Brian; Petersen, Troels; Petit, Elisabeth; Petridis, Andreas; Petridou, Chariclia; Petrolo, Emilio; Petrucci, Fabrizio; Pettersson, Nora Emilia; Pezoa, Raquel; Phillips, Peter William; Piacquadio, Giacinto; Pianori, Elisabetta; Picazio, Attilio; Piccaro, Elisa; Piccinini, Maurizio; Pickering, Mark Andrew; Piegaia, Ricardo; Pignotti, David; Pilcher, James; Pilkington, Andrew; Pina, João Antonio; Pinamonti, Michele; Pinder, Alex; Pinfold, James; Pingel, Almut; Pinto, Belmiro; Pires, Sylvestre; Pitt, Michael; Pizio, Caterina; Plazak, Lukas; Pleier, Marc-Andre; Pleskot, Vojtech; Plotnikova, Elena; Plucinski, Pawel; Pluth, Daniel; Poddar, Sahill; Podlyski, Fabrice; Poettgen, Ruth; Poggioli, Luc; Pohl, David-leon; Pohl, Martin; Polesello, Giacomo; Policicchio, Antonio; Polifka, Richard; Polini, Alessandro; Pollard, Christopher Samuel; Polychronakos, Venetios; Pommès, Kathy; Pontecorvo, Ludovico; Pope, Bernard; Popeneciu, Gabriel Alexandru; Popovic, Dragan; Poppleton, Alan; Pospisil, Stanislav; Potamianos, Karolos; Potrap, Igor; Potter, Christina; Potter, Christopher; Poulard, Gilbert; Poveda, Joaquin; Pozdnyakov, Valery; Pralavorio, Pascal; Pranko, Aliaksandr; Prasad, Srivas; Prell, Soeren; Price, Darren; Price, Joe; Price, Lawrence; Prieur, Damien; Primavera, Margherita; Prince, Sebastien; Proissl, Manuel; Prokofiev, Kirill; Prokoshin, Fedor; Protopapadaki, Eftychia-sofia; Protopopescu, Serban; Proudfoot, James; Przybycien, Mariusz; Przysiezniak, Helenka; Ptacek, Elizabeth; Puddu, Daniele; Pueschel, Elisa; Puldon, David; Purohit, Milind; Puzo, Patrick; Qian, Jianming; Qin, Gang; Qin, Yang; Quadt, Arnulf; Quarrie, David; Quayle, William; Queitsch-Maitland, Michaela; Quilty, Donnchadha; Qureshi, Anum; Radeka, Veljko; Radescu, Voica; Radhakrishnan, Sooraj Krishnan; Radloff, Peter; Rados, Pere; Ragusa, Francesco; Rahal, Ghita; Rajagopalan, Srinivasan; Rammensee, Michael; Rangel-Smith, Camila; Rao, Kanury; Rauscher, Felix; Rave, Stefan; Rave, Tobias Christian; Ravenscroft, Thomas; Raymond, Michel; Read, Alexander Lincoln; Readioff, Nathan Peter; Rebuzzi, Daniela; Redelbach, Andreas; Redlinger, George; Reece, Ryan; Reeves, Kendall; Rehnisch, Laura; Reisin, Hernan; Relich, Matthew; Rembser, Christoph; Ren, Huan; Ren, Zhongliang; Renaud, Adrien; Rescigno, Marco; Resconi, Silvia; Rezanova, Olga; Reznicek, Pavel; Rezvani, Reyhaneh; Richter, Robert; Ridel, Melissa; Rieck, Patrick; Rieger, Julia; Rijssenbeek, Michael; Rimoldi, Adele; Rinaldi, Lorenzo; Ritsch, Elmar; Riu, Imma; Rizatdinova, Flera; Rizvi, Eram; Robertson, Steven; Robichaud-Veronneau, Andree; Robinson, Dave; Robinson, James; Robson, Aidan; Roda, Chiara; Rodrigues, Luis; Roe, Shaun; Røhne, Ole; Rolli, Simona; Romaniouk, Anatoli; Romano, Marino; Romero Adam, Elena; Rompotis, Nikolaos; Ronzani, Manfredi; Roos, Lydia; Ros, Eduardo; Rosati, Stefano; Rosbach, Kilian; Rose, Matthew; Rose, Peyton; Rosendahl, Peter Lundgaard; Rosenthal, Oliver; Rossetti, Valerio; Rossi, Elvira; Rossi, Leonardo Paolo; Rosten, Rachel; Rotaru, Marina; Roth, Itamar; Rothberg, Joseph; Rousseau, David; Royon, Christophe; Rozanov, Alexandre; Rozen, Yoram; Ruan, Xifeng; Rubbo, Francesco; Rubinskiy, Igor; Rud, Viacheslav; Rudolph, Christian; Rudolph, Matthew Scott; Rühr, Frederik; Ruiz-Martinez, Aranzazu; Rurikova, Zuzana; Rusakovich, Nikolai; Ruschke, Alexander; Russell, Heather; Rutherfoord, John; Ruthmann, Nils; Ryabov, Yury; Rybar, Martin; Rybkin, Grigori; Ryder, Nick; Saavedra, Aldo; Sabato, Gabriele; Sacerdoti, Sabrina; Saddique, Asif; Sadrozinski, Hartmut; Sadykov, Renat; Safai Tehrani, Francesco; Sakamoto, Hiroshi; Sakurai, Yuki; Salamanna, Giuseppe; Salamon, Andrea; Saleem, Muhammad; Salek, David; Sales De Bruin, Pedro Henrique; Salihagic, Denis; Salnikov, Andrei; Salt, José; Salvatore, Daniela; Salvatore, Pasquale Fabrizio; Salvucci, Antonio; Salzburger, Andreas; Sampsonidis, Dimitrios; Sanchez, Arturo; Sánchez, Javier; Sanchez Martinez, Victoria; Sandaker, Heidi; Sandbach, Ruth Laura; Sander, Heinz Georg; Sanders, Michiel; Sandhoff, Marisa; Sandoval, Tanya; Sandoval, Carlos; Sandstroem, Rikard; Sankey, Dave; Sansoni, Andrea; Santoni, Claudio; Santonico, Rinaldo; Santos, Helena; Santoyo Castillo, Itzebelt; Sapp, Kevin; Sapronov, Andrey; Saraiva, João; Sarrazin, Bjorn; Sartisohn, Georg; Sasaki, Osamu; Sasaki, Yuichi; Sato, Koji; Sauvage, Gilles; Sauvan, Emmanuel; Savage, Graham; Savard, Pierre; Sawyer, Craig; Sawyer, Lee; Saxon, David; Saxon, James; Sbarra, Carla; Sbrizzi, Antonio; Scanlon, Tim; Scannicchio, Diana; Scarcella, Mark; Scarfone, Valerio; Schaarschmidt, Jana; Schacht, Peter; Schaefer, Douglas; Schaefer, Ralph; Schaepe, Steffen; Schaetzel, Sebastian; Schäfer, Uli; Schaffer, Arthur; Schaile, Dorothee; Schamberger, R~Dean; Scharf, Veit; Schegelsky, Valery; Scheirich, Daniel; Schernau, Michael; Schiavi, Carlo; Schieck, Jochen; Schillo, Christian; Schioppa, Marco; Schlenker, Stefan; Schmidt, Evelyn; Schmieden, Kristof; Schmitt, Christian; Schmitt, Sebastian; Schneider, Basil; Schnellbach, Yan Jie; Schnoor, Ulrike; Schoeffel, Laurent; Schoening, Andre; Schoenrock, Bradley Daniel; Schorlemmer, Andre Lukas; Schott, Matthias; Schouten, Doug; Schovancova, Jaroslava; Schramm, Steven; Schreyer, Manuel; Schroeder, Christian; Schuh, Natascha; Schultens, Martin Johannes; Schultz-Coulon, Hans-Christian; Schulz, Holger; Schumacher, Markus; Schumm, Bruce; Schune, Philippe; Schwanenberger, Christian; Schwartzman, Ariel; Schwarz, Thomas Andrew; Schwegler, Philipp; Schwemling, Philippe; Schwienhorst, Reinhard; Schwindling, Jerome; Schwindt, Thomas; Schwoerer, Maud; Sciacca, Gianfranco; Scifo, Estelle; Sciolla, Gabriella; Scuri, Fabrizio; Scutti, Federico; Searcy, Jacob; Sedov, George; Sedykh, Evgeny; Seema, Pienpen; Seidel, Sally; Seiden, Abraham; Seifert, Frank; Seixas, José; Sekhniaidze, Givi; Sekula, Stephen; Selbach, Karoline Elfriede; Seliverstov, Dmitry; Sellers, Graham; Semprini-Cesari, Nicola; Serfon, Cedric; Serin, Laurent; Serkin, Leonid; Serre, Thomas; Seuster, Rolf; Severini, Horst; Sfiligoj, Tina; Sforza, Federico; Sfyrla, Anna; Shabalina, Elizaveta; Shamim, Mansoora; Shan, Lianyou; Shang, Ruo-yu; Shank, James; Shapiro, Marjorie; Shatalov, Pavel; Shaw, Kate; Shcherbakova, Anna; Shehu, Ciwake Yusufu; Sherwood, Peter; Shi, Liaoshan; Shimizu, Shima; Shimmin, Chase Owen; Shimojima, Makoto; Shiyakova, Mariya; Shmeleva, Alevtina; Shoaleh Saadi, Diane; Shochet, Mel; Shojaii, Seyedruhollah; Short, Daniel; Shrestha, Suyog; Shulga, Evgeny; Shupe, Michael; Shushkevich, Stanislav; Sicho, Petr; Sidiropoulou, Ourania; Sidorov, Dmitri; Sidoti, Antonio; Siegert, Frank; Sijacki, Djordje; Silva, José; Silver, Yiftah; Silverstein, Daniel; Silverstein, Samuel; Simak, Vladislav; Simard, Olivier; Simic, Ljiljana; Simion, Stefan; Simioni, Eduard; Simmons, Brinick; Simon, Dorian; Simoniello, Rosa; Sinervo, Pekka; Sinev, Nikolai; Siragusa, Giovanni; Sircar, Anirvan; Sisakyan, Alexei; Sivoklokov, Serguei; Sjölin, Jörgen; Sjursen, Therese; Skottowe, Hugh Philip; Skubic, Patrick; Slater, Mark; Slavicek, Tomas; Slawinska, Magdalena; Sliwa, Krzysztof; Smakhtin, Vladimir; Smart, Ben; Smestad, Lillian; Smirnov, Sergei; Smirnov, Yury; Smirnova, Lidia; Smirnova, Oxana; Smith, Kenway; Smith, Matthew; Smizanska, Maria; Smolek, Karel; Snesarev, Andrei; Snidero, Giacomo; Snyder, Scott; Sobie, Randall; Socher, Felix; Soffer, Abner; Soh, Dart-yin; Solans, Carlos; Solar, Michael; Solc, Jaroslav; Soldatov, Evgeny; Soldevila, Urmila; Solodkov, Alexander; Soloshenko, Alexei; Solovyanov, Oleg; Solovyev, Victor; Sommer, Philip; Song, Hong Ye; Soni, Nitesh; Sood, Alexander; Sopczak, Andre; Sopko, Bruno; Sopko, Vit; Sorin, Veronica; Sosebee, Mark; Soualah, Rachik; Soueid, Paul; Soukharev, Andrey; South, David; Spagnolo, Stefania; Spanò, Francesco; Spearman, William Robert; Spettel, Fabian; Spighi, Roberto; Spigo, Giancarlo; Spiller, Laurence Anthony; Spousta, Martin; Spreitzer, Teresa; St Denis, Richard Dante; Staerz, Steffen; Stahlman, Jonathan; Stamen, Rainer; Stamm, Soren; Stanecka, Ewa; Stanescu, Cristian; Stanescu-Bellu, Madalina; Stanitzki, Marcel Michael; Stapnes, Steinar; Starchenko, Evgeny; Stark, Jan; Staroba, Pavel; Starovoitov, Pavel; Staszewski, Rafal; Stavina, Pavel; Steinberg, Peter; Stelzer, Bernd; Stelzer, Harald Joerg; Stelzer-Chilton, Oliver; Stenzel, Hasko; Stern, Sebastian; Stewart, Graeme; Stillings, Jan Andre; Stockton, Mark; Stoebe, Michael; Stoicea, Gabriel; Stolte, Philipp; Stonjek, Stefan; Stradling, Alden; Straessner, Arno; Stramaglia, Maria Elena; Strandberg, Jonas; Strandberg, Sara; Strandlie, Are; Strauss, Emanuel; Strauss, Michael; Strizenec, Pavol; Ströhmer, Raimund; Strom, David; Stroynowski, Ryszard; Strubig, Antonia; Stucci, Stefania Antonia; Stugu, Bjarne; Styles, Nicholas Adam; Su, Dong; Su, Jun; Subramaniam, Rajivalochan; Succurro, Antonella; Sugaya, Yorihito; Suhr, Chad; Suk, Michal; Sulin, Vladimir; Sultansoy, Saleh; Sumida, Toshi; Sun, Siyuan; Sun, Xiaohu; Sundermann, Jan Erik; Suruliz, Kerim; Susinno, Giancarlo; Sutton, Mark; Suzuki, Yu; Svatos, Michal; Swedish, Stephen; Swiatlowski, Maximilian; Sykora, Ivan; Sykora, Tomas; Ta, Duc; Taccini, Cecilia; Tackmann, Kerstin; Taenzer, Joe; Taffard, Anyes; Tafirout, Reda; Taiblum, Nimrod; Takai, Helio; Takashima, Ryuichi; Takeda, Hiroshi; Takeshita, Tohru; Takubo, Yosuke; Talby, Mossadek; Talyshev, Alexey; Tam, Jason; Tan, Kong Guan; Tanaka, Junichi; Tanaka, Reisaburo; Tanaka, Satoshi; Tanaka, Shuji; Tanasijczuk, Andres Jorge; Tannenwald, Benjamin Bordy; Tannoury, Nancy; Tapprogge, Stefan; Tarem, Shlomit; Tarrade, Fabien; Tartarelli, Giuseppe Francesco; Tas, Petr; Tasevsky, Marek; Tashiro, Takuya; Tassi, Enrico; Tavares Delgado, Ademar; Tayalati, Yahya; Taylor, Frank; Taylor, Geoffrey; Taylor, Wendy; Teischinger, Florian Alfred; Teixeira Dias Castanheira, Matilde; Teixeira-Dias, Pedro; Temming, Kim Katrin; Ten Kate, Herman; Teng, Ping-Kun; Teoh, Jia Jian; Tepel, Fabian-Phillipp; Terada, Susumu; Terashi, Koji; Terron, Juan; Terzo, Stefano; Testa, Marianna; Teuscher, Richard; Therhaag, Jan; Theveneaux-Pelzer, Timothée; Thomas, Juergen; Thomas-Wilsker, Joshuha; Thompson, Emily; Thompson, Paul; Thompson, Ray; Thompson, Stan; Thomsen, Lotte Ansgaard; Thomson, Evelyn; Thomson, Mark; Thong, Wai Meng; Thun, Rudolf; Tian, Feng; Tibbetts, Mark James; Tikhomirov, Vladimir; Tikhonov, Yury; Timoshenko, Sergey; Tiouchichine, Elodie; Tipton, Paul; Tisserant, Sylvain; Todorov, Theodore; Todorova-Nova, Sharka; Tojo, Junji; Tokár, Stanislav; Tokushuku, Katsuo; Tollefson, Kirsten; Tolley, Emma; Tomlinson, Lee; Tomoto, Makoto; Tompkins, Lauren; Toms, Konstantin; Topilin, Nikolai; Torrence, Eric; Torres, Heberth; Torró Pastor, Emma; Toth, Jozsef; Touchard, Francois; Tovey, Daniel; Tran, Huong Lan; Trefzger, Thomas; Tremblet, Louis; Tricoli, Alessandro; Trigger, Isabel Marian; Trincaz-Duvoid, Sophie; Tripiana, Martin; Trischuk, William; Trocmé, Benjamin; Troncon, Clara; Trottier-McDonald, Michel; Trovatelli, Monica; True, Patrick; Trzebinski, Maciej; Trzupek, Adam; Tsarouchas, Charilaos; Tseng, Jeffrey; Tsiareshka, Pavel; Tsionou, Dimitra; Tsipolitis, Georgios; Tsirintanis, Nikolaos; Tsiskaridze, Shota; Tsiskaridze, Vakhtang; Tskhadadze, Edisher; Tsukerman, Ilya; Tsulaia, Vakhtang; Tsuno, Soshi; Tsybychev, Dmitri; Tudorache, Alexandra; Tudorache, Valentina; Tuna, Alexander Naip; Tupputi, Salvatore; Turchikhin, Semen; Turecek, Daniel; Turk Cakir, Ilkay; Turra, Ruggero; Turvey, Andrew John; Tuts, Michael; Tykhonov, Andrii; Tylmad, Maja; Tyndel, Mike; Ueda, Ikuo; Ueno, Ryuichi; Ughetto, Michael; Ugland, Maren; Uhlenbrock, Mathias; Ukegawa, Fumihiko; Unal, Guillaume; Undrus, Alexander; Unel, Gokhan; Ungaro, Francesca; Unno, Yoshinobu; Unverdorben, Christopher; Urban, Jozef; Urbaniec, Dustin; Urquijo, Phillip; Usai, Giulio; Usanova, Anna; Vacavant, Laurent; Vacek, Vaclav; Vachon, Brigitte; Valencic, Nika; Valentinetti, Sara; Valero, Alberto; Valery, Loic; Valkar, Stefan; Valladolid Gallego, Eva; Vallecorsa, Sofia; Valls Ferrer, Juan Antonio; Van Den Wollenberg, Wouter; Van Der Deijl, Pieter; van der Geer, Rogier; van der Graaf, Harry; Van Der Leeuw, Robin; van der Ster, Daniel; van Eldik, Niels; van Gemmeren, Peter; Van Nieuwkoop, Jacobus; van Vulpen, Ivo; van Woerden, Marius Cornelis; Vanadia, Marco; Vandelli, Wainer; Vanguri, Rami; Vaniachine, Alexandre; Vannucci, Francois; Vardanyan, Gagik; Vari, Riccardo; Varnes, Erich; Varol, Tulin; Varouchas, Dimitris; Vartapetian, Armen; Varvell, Kevin; Vazeille, Francois; Vazquez Schroeder, Tamara; Veatch, Jason; Veloso, Filipe; Velz, Thomas; Veneziano, Stefano; Ventura, Andrea; Ventura, Daniel; Venturi, Manuela; Venturi, Nicola; Venturini, Alessio; Vercesi, Valerio; Verducci, Monica; Verkerke, Wouter; Vermeulen, Jos; Vest, Anja; Vetterli, Michel; Viazlo, Oleksandr; Vichou, Irene; Vickey, Trevor; Vickey Boeriu, Oana Elena; Viehhauser, Georg; Viel, Simon; Vigne, Ralph; Villa, Mauro; Villaplana Perez, Miguel; Vilucchi, Elisabetta; Vincter, Manuella; Vinogradov, Vladimir; Virzi, Joseph; Vivarelli, Iacopo; Vives Vaque, Francesc; Vlachos, Sotirios; Vladoiu, Dan; Vlasak, Michal; Vogel, Adrian; Vogel, Marcelo; Vokac, Petr; Volpi, Guido; Volpi, Matteo; von der Schmitt, Hans; von Radziewski, Holger; von Toerne, Eckhard; Vorobel, Vit; Vorobev, Konstantin; Vos, Marcel; Voss, Rudiger; Vossebeld, Joost; Vranjes, Nenad; Vranjes Milosavljevic, Marija; Vrba, Vaclav; Vreeswijk, Marcel; Vu Anh, Tuan; Vuillermet, Raphael; Vukotic, Ilija; Vykydal, Zdenek; Wagner, Peter; Wagner, Wolfgang; Wahlberg, Hernan; Wahrmund, Sebastian; Wakabayashi, Jun; Walder, James; Walker, Rodney; Walkowiak, Wolfgang; Wall, Richard; Waller, Peter; Walsh, Brian; Wang, Chao; Wang, Chiho; Wang, Fuquan; Wang, Haichen; Wang, Hulin; Wang, Jike; Wang, Jin; Wang, Kuhan; Wang, Rui; Wang, Song-Ming; Wang, Tan; Wang, Xiaoxiao; Wanotayaroj, Chaowaroj; Warburton, Andreas; Ward, Patricia; Wardrope, David Robert; Warsinsky, Markus; Washbrook, Andrew; Wasicki, Christoph; Watkins, Peter; Watson, Alan; Watson, Ian; Watson, Miriam; Watts, Gordon; Watts, Stephen; Waugh, Ben; Webb, Samuel; Weber, Michele; Weber, Stefan Wolf; Webster, Jordan S; Weidberg, Anthony; Weinert, Benjamin; Weingarten, Jens; Weiser, Christian; Weits, Hartger; Wells, Phillippa; Wenaus, Torre; Wendland, Dennis; Weng, Zhili; Wengler, Thorsten; Wenig, Siegfried; Wermes, Norbert; Werner, Matthias; Werner, Per; Wessels, Martin; Wetter, Jeffrey; Whalen, Kathleen; White, Andrew; White, Martin; White, Ryan; White, Sebastian; Whiteson, Daniel; Wicke, Daniel; Wickens, Fred; Wiedenmann, Werner; Wielers, Monika; Wienemann, Peter; Wiglesworth, Craig; Wiik-Fuchs, Liv Antje Mari; Wijeratne, Peter Alexander; Wildauer, Andreas; Wildt, Martin Andre; Wilkens, Henric George; Williams, Hugh; Williams, Sarah; Willis, Christopher; Willocq, Stephane; Wilson, Alan; Wilson, John; Wingerter-Seez, Isabelle; Winklmeier, Frank; Winter, Benedict Tobias; Wittgen, Matthias; Wittkowski, Josephine; Wollstadt, Simon Jakob; Wolter, Marcin Wladyslaw; Wolters, Helmut; Wosiek, Barbara; Wotschack, Jorg; Woudstra, Martin; Wozniak, Krzysztof; Wright, Michael; Wu, Mengqing; Wu, Sau Lan; Wu, Xin; Wu, Yusheng; Wyatt, Terry Richard; Wynne, Benjamin; Xella, Stefania; Xiao, Meng; Xu, Da; Xu, Lailin; Yabsley, Bruce; Yacoob, Sahal; Yakabe, Ryota; Yamada, Miho; Yamaguchi, Hiroshi; Yamaguchi, Yohei; Yamamoto, Akira; Yamamoto, Shimpei; Yamamura, Taiki; Yamanaka, Takashi; Yamauchi, Katsuya; Yamazaki, Yuji; Yan, Zhen; Yang, Haijun; Yang, Hongtao; Yang, Yi; Yanush, Serguei; Yao, Liwen; Yao, Weiming; Yasu, Yoshiji; Yatsenko, Elena; Yau Wong, Kaven Henry; Ye, Jingbo; Ye, Shuwei; Yeletskikh, Ivan; Yen, Andy L; Yildirim, Eda; Yilmaz, Metin; Yorita, Kohei; Yoshida, Rikutaro; Yoshihara, Keisuke; Young, Charles; Young, Christopher John; Youssef, Saul; Yu, David Ren-Hwa; Yu, Jaehoon; Yu, Jiaming; Yu, Jie; Yuan, Li; Yurkewicz, Adam; Yusuff, Imran; Zabinski, Bartlomiej; Zaidan, Remi; Zaitsev, Alexander; Zaman, Aungshuman; Zambito, Stefano; Zanello, Lucia; Zanzi, Daniele; Zeitnitz, Christian; Zeman, Martin; Zemla, Andrzej; Zengel, Keith; Zenin, Oleg; Ženiš, Tibor; Zerwas, Dirk; Zevi della Porta, Giovanni; Zhang, Dongliang; Zhang, Fangzhou; Zhang, Huaqiao; Zhang, Jinlong; Zhang, Lei; Zhang, Ruiqi; Zhang, Xueyao; Zhang, Zhiqing; Zhao, Xiandong; Zhao, Yongke; Zhao, Zhengguo; Zhemchugov, Alexey; Zhong, Jiahang; Zhou, Bing; Zhou, Chen; Zhou, Lei; Zhou, Li; Zhou, Ning; Zhu, Cheng Guang; Zhu, Hongbo; Zhu, Junjie; Zhu, Yingchun; Zhuang, Xuai; Zhukov, Konstantin; Zibell, Andre; Zieminska, Daria; Zimine, Nikolai; Zimmermann, Christoph; Zimmermann, Robert; Zimmermann, Simone; Zimmermann, Stephanie; Zinonos, Zinonas; Ziolkowski, Michael; Zobernig, Georg; Zoccoli, Antonio; zur Nedden, Martin; Zurzolo, Giovanni; Zwalinski, Lukasz

    2015-07-16

    We report the observation of Higgs boson decays to $WW^{\\ast}$ based on an excess over background of 6.1 standard deviations in the dilepton final state, where the Standard Model expectation is 5.8 standard deviations. Evidence for the vector-boson fusion (VBF) production process is obtained with a significance of 3.2 standard deviations. The results are obtained from a data sample corresponding to an integrated luminosity of $25~\\textrm{pb}^{-1}$ from $\\sqrt{s}=7$ and 8 TeV $pp$ collisions recorded by the ATLAS detector at the LHC. For a Higgs boson mass of 125.36 GeV, the ratio of the measured value to the expected value of the total production cross section times branching fraction is $1.09^{+0.16}_{-0.15}~\\textrm{(stat.)}^{+0.17}_{-0.14}~\\textrm{(syst.)}$. The corresponding ratios for the gluon fusion and vector-boson fusion production mechanisms are $1.02\\pm 0.19~\\textrm{(stat.)}^{+0.22}_{-0.18}~\\textrm{(syst.)}$ and $1.27^{+0.44}_{-0.40}~\\textrm{(stat.)}^{+0.30}_{-0.21}~\\textrm{(syst.)}$, respectively. ...

  6. Perspectives on reactor safety. Revision 1

    International Nuclear Information System (INIS)

    Haskin, F.E.; Hodge, S.A.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  7. Perspectives on reactor safety. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  8. Passive safety systems for integral reactors

    International Nuclear Information System (INIS)

    Kuul, V.S.; Samoilov, O.B.

    1996-01-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs

  9. Passive safety systems for integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuul, V S; Samoilov, O B [OKB Mechanical Engineering (Russian Federation)

    1996-12-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs.

  10. Expressão do Mg+2, CK, AST e LDH em equinos finalistas de provas de enduro Endurance horses finalists: expression of Mg+2, CK, AST and LDH in horse finalists of endurance race

    Directory of Open Access Journals (Sweden)

    Juliana V.F. Sales

    2013-01-01

    Full Text Available Nos últimos anos, o equino atleta vem sendo cada vez mais requerido. Dessa forma, as exigências por alto desempenho têm fomentado o interesse pelo estudo das afecções relacionadas com a fisiopatologia de diversas enfermidades dos equinos. A relação entre o íon magnésio e o exercício físico tem recebido atenção significativa visto que este íon está intimamente relacionado ao tecido muscular estriado esquelético. Além disso, dentre as principais estratégias para a detecção e acompanhamento clínico de lesões musculares, destacam-se a avaliação das atividades das enzimas creatino quinase (CK, lactato desidrogenase (LDH e aspartato aminotransferase (AST. A busca pelo estabelecimento de parâmetros que se relacionam entre si é um fator determinante na compreensão de alterações fisiológicas encontradas diante do esforço em equinos atletas. Desta forma, o presente trabalho teve como objetivo determinar como as concentrações sanguíneas do íon magnésio e as atividades enzimáticas das enzimas CK, LDH e AST comportaram-se em equinos Puro Sangue Árabe finalistas de provas de enduro de 90km e relacionar as possíveis alterações com o tipo de esforço físico desempenhado pelos animais. Foram avaliadas a atividade enzimática das enzimas CK, LDH, AST e a concentração do íon magnésio no exercício em relação ao repouso de 14 equinos clinicamente hígidos da raça Puro Sangue Árabe, sendo 9 machos e 5 fêmeas, com idades variando entre 6 a 12 anos, submetidos a treinamento para enduro e participantes de provas de 90 km. Pode-se observar que as variáveis acima mencionadas sofreram aumento com diferença estatística em relação ao repouso. O exercício físico de enduro determinou a ocorrência de alterações nas atividades enzimáticas das enzimas CK (p≤0,001, LDH (p=0,0001, AST (p=0,0007 e na concentração do íon magnésio (p=0,0004, no exercício em relação ao repouso (p≤0,05. Fato que determinou altera

  11. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  12. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  13. Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

    International Nuclear Information System (INIS)

    Perry, E.F.

    1995-01-01

    The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

  14. Reactor utilization, Part 1

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1981-01-01

    The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

  15. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Salahuddin, A.; Israr, M.; Hussain, M.

    2002-01-01

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  16. RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1985-01-01

    According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

  17. New instrumentation for the IPR-R1 reactor of CDTN

    International Nuclear Information System (INIS)

    Carvalho, P.V.R. de.

    1992-01-01

    The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

  18. 76 FR 51459 - Office of Commercial Space Transportation (AST); Notice of Availability of the Record of Decision...

    Science.gov (United States)

    2011-08-18

    ... Action, FAA/AST could issue, renew, or modify launch operator licenses for Atlas V and Delta IV... the Evolved Expendable Launch Vehicle (EELV) Program, Which Include Atlas V and Delta IV Vehicles, From Cape Canaveral Air Force Station (CCAFS), Florida and Vandenberg Air Force Base (VAFB), California...

  19. Evidence for the Standard Model Higgs boson in the WW$^{\\ast}$ decay mode using the data collected by the ATLAS detector at the LHC

    CERN Document Server

    Jovicevic, Jelena; Strandberg, Jonas

    2015-11-20

    The $H \\rightarrow WW^{\\ast}$ channel was one of the three search channels contributing in the observation of the Higgs boson at the ATLAS experiment in July 2012. Nowadays, this channel represents an important ingredient in the determination of the couplings and properties of the newly discovered particle. This thesis reports the search for and observation of the Higgs boson in the $H \\rightarrow WW^{\\ast}$ decay mode using the ATLAS detector at the CERN Large Hadron Collider. The analysis of events in which the Higgs boson is produced in the gluon-gluon fusion process and is associated with no more than one jet, is outlined in detail. The datasets used are the proton-proton collisions collected with the ATLAS detector at a centre of mass energy of 8 TeV during 2012 and 7 TeV during 2011, corresponding to a total integrated luminosity of 25 fb$^{-1}$. An excess over the predicted number of background events is observed in the data. The significance of the excess is estimated to be 6.1 standard deviations, ...

  20. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  1. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

    International Nuclear Information System (INIS)

    Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

    2002-01-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  2. New human machine interface for VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

    2002-01-01

    The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

  3. Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

    International Nuclear Information System (INIS)

    1988-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

  4. Department of Reactor Technology annual progress report 1 January -31 December 1977

    International Nuclear Information System (INIS)

    1978-04-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

  5. Lepton-flavor universality violation in R K and {R}_{D{_{(\\ast )}}} from warped space

    Science.gov (United States)

    Megías, Eugenio; Quirós, Mariano; Salas, Lindber

    2017-07-01

    Some anomalies in the processes b → sℓℓ ( ℓ = μ, e) and b\\to cℓ {\\overline{ν}}_{ℓ } ( ℓ = τ, μ, e), in particular in the observables R K and {R}_{D{_{(\\ast )}}} , have been found by the BaBar, LHCb and Belle collaborations, leading to a possible lepton flavor universality violation. If these anomalies were confirmed they would inevitably lead to physics beyond the Standard Model. In this paper we try to accommodate the present anomalies in an extra dimensional theory, solving the naturalness problem of the Standard Model by means of a warped metric with a strong conformality violation near the infra-red brane. The R K anomaly can be accommodated provided that the left-handed bottom quark and muon lepton have some degree of compositeness in the dual theory. The theory is consistent with all electroweak and flavor observables, and with all direct searches of Kaluza-Klein electroweak gauge bosons and gluons. The fermion spectrum, and fermion mixing angles, can be reproduced by mostly elementary right-handed bottom quarks, and tau and muon leptons. Moreover the {R}_{D{_{(\\ast )}}} anomaly requires a strong degree of compositeness for the left-handed tau leptons, which turns out to be in tension with experimental data on the {g}_{τ_L}^Z coupling, possibly unless some degree of fine-tuning is introduced in the fixing of the CKM matrix.

  6. Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

    International Nuclear Information System (INIS)

    1989-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  7. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Ratanatongchai, Wichian

    2009-07-01

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  8. New measuring and protection system at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Jurickova, M.

    2006-01-01

    The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

  9. Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

    International Nuclear Information System (INIS)

    1991-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  10. Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

    International Nuclear Information System (INIS)

    1992-08-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  11. Thermohydraulic analysis for power increase of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Umbehaun, Pedro E.; Bastos, Jose L.F.

    1996-01-01

    In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

  12. Decommissioning and decontrolling the R1-reactor

    International Nuclear Information System (INIS)

    Bergman, C.; Holmberg, B.T.

    1985-01-01

    Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

  13. AST Critical Propulsion and Noise Reduction Technologies for Future Commercial Subsonic Engines Area of Interest 1.0: Reliable and Affordable Control Systems

    Science.gov (United States)

    Myers, William; Winter, Steve

    2006-01-01

    The General Electric Reliable and Affordable Controls effort under the NASA Advanced Subsonic Technology (AST) Program has designed, fabricated, and tested advanced controls hardware and software to reduce emissions and improve engine safety and reliability. The original effort consisted of four elements: 1) a Hydraulic Multiplexer; 2) Active Combustor Control; 3) a Variable Displacement Vane Pump (VDVP); and 4) Intelligent Engine Control. The VDVP and Intelligent Engine Control elements were cancelled due to funding constraints and are reported here only to the state they progressed. The Hydraulic Multiplexing element developed and tested a prototype which improves reliability by combining the functionality of up to 16 solenoids and servo-valves into one component with a single electrically powered force motor. The Active Combustor Control element developed intelligent staging and control strategies for low emission combustors. This included development and tests of a Controlled Pressure Fuel Nozzle for fuel sequencing, a Fuel Multiplexer for individual fuel cup metering, and model-based control logic. Both the Hydraulic Multiplexer and Controlled Pressure Fuel Nozzle system were cleared for engine test. The Fuel Multiplexer was cleared for combustor rig test which must be followed by an engine test to achieve full maturation.

  14. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  15. Truth-telling contra perfectionist liberalism: Muslim parrhēsíastes in Denmark

    DEFF Research Database (Denmark)

    Renders, Johannes

    In this paper, I first offer a general outline and reflection on the notion of parrhēsía (truth-telling), as popularized by Foucault. Secondly, I discuss Foucault’s history of problematizations, with comments on what he called “games of truth” and the Cartesian conception of truth-telling. Thirdly......, I sketch a trend in the current Danish public and political sphere, defining the notion of “perfectionist liberalism” and how it translates to the Danish context, including concrete examples and notes on “liberal intolerance arguments”. Lastly, I address the condition of Muslim parrhēsíastes (truth...

  16. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

    1981-12-01

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  17. Department of Reactor Technology: annual progress report 1 January - 31 December 1976

    International Nuclear Information System (INIS)

    1977-06-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

  18. Un cinéaste antifasciste à Paris : Slatan Dudow (1934-1939)

    OpenAIRE

    Trugeon, Mélanie

    2011-01-01

    Réfugié à Paris après l’arrivée de Hitler au pouvoir en Allemagne, le metteur en scène de théâtre et cinéaste bulgare Slatan Dudow poursuit son travail militant sur scène, à la radio et sur les écrans dans la capitale française jusqu’à son expulsion en 1939-1940 où il trouve alors refuge en Suisse. À partir d’un riche fonds d’archives déposé aux Archives Nationales, inexploité à ce jour, cet article retrace les conditions et les aléas de cette activité artistique et politique, en particulier ...

  19. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Maretti Junior, F.

    1980-01-01

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  20. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  1. Department of Reactor Technology annual progress report 1 January - 31 December 1978

    International Nuclear Information System (INIS)

    1979-04-01

    The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  2. Effect of two kinds of porcelain crown on AST, ALP, TNF-α, IL-8, GP-x and MDA levels in gingival crevicular fluid

    Directory of Open Access Journals (Sweden)

    Ya-Ling Wang

    2016-09-01

    Full Text Available Objective: To investigate the effect of two kinds of porcelain crown on AST, ALP, TNF-α, IL-8, GP-x and MDA levels in gingival crevicular fluid. Methods: A total of 80 patients with dental porcelain crowns at front teeth during February 2013 to February 2016 were randomly divided into cobalt-chromium alloy PFM group (n=40 and gold alloy PFM group (n=40. After 6 months, the amount of gingival crevicular fluid, GI, PD, AST, ALP, TNF-α, IL-8, GP-x and MDA levels in gingival crevicular fluid were recorded and analyzed. Results: There were no differences in amount of gingival crevicular fluid, GI and PD before treatment of the two groups (P>0.05. After treatment, the amount of gingival crevicular fluid, GI and PD of the two groups were significantly higher than before treatment (P0.05. After treatment, the AST, ALP, TNF-α, IL-8 and MDA levels in gingival crevicular fluid of the two groups were significantly higher than before treatment (P<0.05, but that of the gold alloy PFM group were significantly lower than cobalt-chromium alloy PFM group (P<0.05. After treatment, the GP-x level in gingival crevicular fluid of the two groups were significantly lower than before treatment (P<0.05, but that of the gold alloy PFM group were significantly higher than cobalt-chromium alloy PFM group (P<0.05. Conclusions: Gold alloy PFM can significantly reduce the AST, ALP, TNF-α, IL-8 and MDA levels in gingival crevicular fluid, improve the GP-x level in gingival crevicular fluid, shows better biocompatibility and clinical outcomes than cobalt-chromium alloy PFM.

  3. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    International Nuclear Information System (INIS)

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  4. Reactor engineering department annual report. April 1, 1993-March 31, 1994

    International Nuclear Information System (INIS)

    1994-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

  5. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  6. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    International Nuclear Information System (INIS)

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  7. Modification of the IAN-R1 reactor

    International Nuclear Information System (INIS)

    Jaime, J.; Ahumada, S.; Spin, R.A.

    1990-01-01

    The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

  8. Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Kassar, E.

    1986-01-01

    The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

  9. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  10. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    International Nuclear Information System (INIS)

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  11. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

  12. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  13. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  14. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  15. Present and future activities of TRIGA RC-1 Reactor

    International Nuclear Information System (INIS)

    Festinesi, A.

    1986-01-01

    A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

  16. Neutronic studies in the enrichment reduction of research reactor IEAR-1

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Fanaro, L.C.B.; Mai, L.A.; Ferreira, P.S.B.; Garone, J.G.M.

    1987-01-01

    In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)

  17. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  18. Reliabitity study of the accumulator system for Angra-1 reactor

    International Nuclear Information System (INIS)

    Santos Maciel, C.C.R.

    1980-01-01

    The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

  19. Topotactic conversion of β-helix-layered silicate into AST-type zeolite through successive interlayer modifications.

    Science.gov (United States)

    Asakura, Yusuke; Takayama, Ryosuke; Shibue, Toshimichi; Kuroda, Kazuyuki

    2014-02-10

    AST-type zeolite with a plate morphology can be synthesized by topotactic conversion of a layered silicate (β-helix-layered silicate; HLS) by using N,N-dimethylpropionamide (DPA) to control the layer stacking of silicate layers and the subsequent interlayer condensation. Treatment of HLS twice with 1) hydrochloric acid/ethanol and 2) dimethylsulfoxide (DMSO) are needed to remove interlayer hydrated Na ions and tetramethylammonium (TMA) ions in intralayer cup-like cavities (intracavity TMA ions), both of which are introduced during the preparation of HLS. The utilization of an amide molecule is effective for the control of the stacking sequence of silicate layers. This method could be applicable to various layered silicates that cannot be topotactically converted into three-dimensional networks by simple interlayer condensation by judicious choice of amide molecules. Copyright © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  20. Design of a new research reactor : 1st year conceptual design

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T.

    2004-01-01

    A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

  1. Development, Dedication and Application of an Automatic Seismic Trip System for Nuclear Power Plants of Taiwan Power Company

    International Nuclear Information System (INIS)

    Liao, Hsin-kai; Lee, Chung-lin; Chen, Chang-kuo; Hsu, Yao-tung; Shyu, Shian-shing

    2011-01-01

    This paper describes the setups of Automatic Seismic Trip System (ASTS), including development, dedication and implementation, for Nuclear Power Plants (NPPs) of Taiwan Power Company (TPC). The purposed ASTS was designed to trip the reactor when big earthquake occurs. These ASTS were classified as class 1E equipment. They were developed and dedicated for safety applications in accordance with IEEE 323-1983, IEEE 344-1987, IEEE 383-1974 and Reg. Guide 1.180 R1. In order to meet the technical specification required by TPC, three sub-units in the ASTS were developed: Earthquake sensors: Kinemetrices FBA-23 triaxial accelerometers are selected since they were successfully used in Taiwan for seismic monitoring for more than 10 years. Signal conditioning module: It is designed to reduce noise from motion accelerometer (FBA-23) and then transmit seismic signal to the set-point and trip unit via instrument amplify circuit, 0.1 to 10Hz band pass filter circuit, absolute-value converter and voltage to current converter. Trip control module: after comparing the seismic signal level and set-point, the result will decide whether to drive the output relay or not. The output relay is used as the interface between ASTS and the reactor protection system in NPP. For the commercial grade item dedication for safety application, five processes were conducted. Those processes are Seismic test: to use plant specific required response spectrum (RRS), the test required spectrum should envelop RRS: Seismic auto-trip accuracy test: must not trip when filtered PA below set point minus 0.05g, and must trip when filtered PA exceeds set point over 0.05g. Trip signals occurred within 10 second interval are considered as same events: NEMA4 water proof test for sensor box: Anti-radiation test: 8.76x100 rads over 40 years: EMI/EMC test: follow RG 1.180 requirement. The ASTS were installed in three NPPs, six units in total, without connection to RPS in 2006. After one year reliable operation, the

  2. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    Matejka, K.

    1993-01-01

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

  3. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    Petit, A.; Brie, M.

    1996-01-01

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  4. Sandia reactor kinetics codes: SAK and PK1D

    International Nuclear Information System (INIS)

    Pickard, P.S.; Odom, J.P.

    1978-01-01

    The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

  5. Implementation and evaluation of the 5As framework of obesity management in primary care: design of the 5As Team (5AsT) randomized control trial.

    Science.gov (United States)

    Campbell-Scherer, Denise L; Asselin, Jodie; Osunlana, Adedayo M; Fielding, Sheri; Anderson, Robin; Rueda-Clausen, Christian F; Johnson, Jeffrey A; Ogunleye, Ayodele A; Cave, Andrew; Manca, Donna; Sharma, Arya M

    2014-06-19

    Obesity is a pressing public health concern, which frequently presents in primary care. With the explosive obesity epidemic, there is an urgent need to maximize effective management in primary care. The 5As of Obesity Management™ (5As) are a collection of knowledge tools developed by the Canadian Obesity Network. Low rates of obesity management visits in primary care suggest provider behaviour may be an important variable. The goal of the present study is to increase frequency and quality of obesity management in primary care using the 5As Team (5AsT) intervention to change provider behaviour. The 5AsT trial is a theoretically informed, pragmatic randomized controlled trial with mixed methods evaluation. Clinic-based multidisciplinary teams (RN/NP, mental health, dietitians) will be randomized to control or the 5AsT intervention group, to participate in biweekly learning collaborative sessions supported by internal and external practice facilitation. The learning collaborative content addresses provider-identified barriers to effective obesity management in primary care. Evidence-based shared decision making tools will be co-developed and iteratively tested by practitioners. Evaluation will be informed by the RE-AIM framework. The primary outcome measure, to which participants are blinded, is number of weight management visits/full-time equivalent (FTE) position. Patient-level outcomes will also be assessed, through a longitudinal cohort study of patients from randomized practices. Patient outcomes include clinical (e.g., body mass index [BMI], blood pressure), health-related quality of life (SF-12, EQ5D), and satisfaction with care. Qualitative data collected from providers and patients will be evaluated using thematic analysis to understand the context, implementation and effectiveness of the 5AsT program. The 5AsT trial will provide a wide range of insights into current practices, knowledge gaps and barriers that limit obesity management in primary practice

  6. Ageing problems and renovation programme of ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Khattab, M.S.; Sultan, M.A.

    1995-01-01

    Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

  7. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  8. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

  9. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1987-01-01

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

  10. Tetranectin Binds to the Kringle 1-4 Form of Angiostatin and Modifies Its Functional Activity

    DEFF Research Database (Denmark)

    Mogues, Tirsit; Etzerodt, Michael; Hall, Crystal

    2004-01-01

    influence cancer progression is by altering activities of plasminogen or the plasminogen fragment, angiostatin. Tetranectin was found to bind to the kringle 1-4 form of angiostatin (AST $;{\\text{K1-4}}$ ). In addition, tetranectin inhibited binding of plasminogen or AST $;{\\text{K1-4}}$ to extracellular...... matrix (ECM) deposited by endothelial cells. Finally, tetranectin partially counteracted the ability of AST $;{\\text{K1-4}}$ to inhibit proliferation of endothelial cells. This latter effect of tetranectin was specific for AST $;{\\text{K1-4}}$ since it did not counteract the antiproliferative activities...

  11. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Kuruc, J.; Matel, L.

    2007-01-01

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

  12. Demolition of the FRJ-1 research reactor (MERLIN)

    International Nuclear Information System (INIS)

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-01-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

  13. Evaporation rate measurement in the pool of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

    2000-01-01

    The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

  14. New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

    Energy Technology Data Exchange (ETDEWEB)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  15. Measurements of the S-wave fraction in $B^{0}\\rightarrow K^{+}\\pi^{-}\\mu^{+}\\mu^{-}$ decays and the $B^{0}\\rightarrow K^{\\ast}(892)^{0}\\mu^{+}\\mu^{-}$ differential branching fraction

    CERN Document Server

    Aaij, Roel; Adinolfi, Marco; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Andreassi, Guido; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; d'Argent, Philippe; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Bel, Lennaert; Bellee, Violaine; Belloli, Nicoletta; Belous, Konstantin; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bertolin, Alessandro; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bifani, Simone; Billoir, Pierre; Bird, Thomas; Birnkraut, Alex; Bitadze, Alexander; Bizzeti, Andrea; Blake, Thomas; Blanc, Frederic; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Boettcher, Thomas; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borisyak, Maxim; Borsato, Martino; Bossu, Francesco; Boubdir, Meriem; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Braun, Svende; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Buchanan, Emma; Burr, Christopher; Bursche, Albert; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Capriotti, Lorenzo; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carniti, Paolo; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cavallero, Giovanni; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chatzikonstantinidis, Georgios; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chobanova, Veronika; Chrzaszcz, Marcin; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cogoni, Violetta; Cojocariu, Lucian; Collazuol, Gianmaria; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Crocombe, Andrew; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dall'Occo, Elena; Dalseno, Jeremy; David, Pieter; Davis, Adam; De Aguiar Francisco, Oscar; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Simone, Patrizia; Dean, Cameron Thomas; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Demmer, Moritz; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Dey, Biplab; Di Canto, Angelo; Dijkstra, Hans; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dovbnya, Anatoliy; Dreimanis, Karlis; Dufour, Laurent; Dujany, Giulio; Dungs, Kevin; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Déléage, Nicolas; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferrari, Fabio; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fleuret, Frederic; Fohl, Klaus; Fontana, Marianna; Fontanelli, Flavio; Forshaw, Dean Charles; Forty, Roger; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Färber, Christian; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garra Tico, Jordi; Garrido, Lluis; Garsed, Philip John; Gascon, David; Gaspar, Clara; Gavardi, Laura; Gazzoni, Giulio; Gerick, David; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianì, Sebastiana; Gibson, Valerie; Girard, Olivier Göran; Giubega, Lavinia-Helena; Gizdov, Konstantin; Gligorov, V.V.; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gorelov, Igor Vladimirovich; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graverini, Elena; Graziani, Giacomo; Grecu, Alexandru; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Göbel, Carla; Hadavizadeh, Thomas; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heister, Arno; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hombach, Christoph; Hulsbergen, Wouter; Humair, Thibaud; Hushchyn, Mikhail; Hussain, Nazim; Hutchcroft, David; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jawahery, Abolhassan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kecke, Matthieu; Kelsey, Matthew; Kenyon, Ian; Kenzie, Matthew; Ketel, Tjeerd; Khairullin, Egor; Khanji, Basem; Khurewathanakul, Chitsanu; Kirn, Thomas; Klaver, Suzanne; Klimaszewski, Konrad; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Kozachuk, Anastasiia; Kozeiha, Mohamad; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krokovny, Pavel; Kruse, Florian; Krzemien, Wojciech; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kuonen, Axel Kevin; Kurek, Krzysztof; Kvaratskheliya, Tengiz; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Leflat, Alexander; Lefrançois, Jacques; Lefèvre, Regis; Lemaitre, Florian; Lemos Cid, Edgar; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Liu, Xuesong; Loh, David; Longstaff, Iain; Lopes, Jose; Lucchesi, Donatella; Lucio Martinez, Miriam; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Lusiani, Alberto; Lyu, Xiao-Rui; Machefert, Frederic; Maciuc, Florin; Maev, Oleg; Maguire, Kevin; Malde, Sneha; Malinin, Alexander; Maltsev, Timofei; Manca, Giulia; Mancinelli, Giampiero; Manning, Peter Michael; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Marks, Jörg; Martellotti, Giuseppe; Martin, Morgan; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massacrier, Laure Marie; Massafferri, André; Matev, Rosen; Mathad, Abhijit; Mathe, Zoltan; Matteuzzi, Clara; Mauri, Andrea; Maurin, Brice; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; Meadows, Brian; Meier, Frank; Meissner, Marco; Melnychuk, Dmytro; Merk, Marcel; Michielin, Emanuele; Milanes, Diego Alejandro; Minard, Marie-Noelle; Mitzel, Dominik Stefan; Molina Rodriguez, Josue; Monroy, Ignacio Alberto; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Mulder, Mick; Mussini, Manuel; Müller, Dominik; Müller, Janine; Müller, Katharina; Müller, Vanessa; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nandi, Anita; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen-Mau, Chung; Niess, Valentin; Nieswand, Simon; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Ogilvy, Stephen; Oldeman, Rudolf; Onderwater, Gerco; Otalora Goicochea, Juan Martin; Otto, Adam; Owen, Patrick; Oyanguren, Maria Aranzazu; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Pappenheimer, Cheryl; Parker, William; Parkes, Christopher; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pearce, Alex; Pellegrino, Antonio; Penso, Gianni; Pepe Altarelli, Monica; Perazzini, Stefano; Perret, Pascal; Pescatore, Luca; Petridis, Konstantinos; Petrolini, Alessandro; Petrov, Aleksandr; Petruzzo, Marco; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pikies, Malgorzata; Pinci, Davide; Pistone, Alessandro; Piucci, Alessio; Playfer, Stephen; Plo Casasus, Maximo; Poikela, Tuomas; Polci, Francesco; Poluektov, Anton; Polyakov, Ivan; Polycarpo, Erica; Pomery, Gabriela Johanna; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Quagliani, Renato; Rachwal, Bartolomiej; Rademacker, Jonas; Rama, Matteo; Ramos Pernas, Miguel; Rangel, Murilo; Raniuk, Iurii; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; dos Reis, Alberto; Remon Alepuz, Clara; Renaudin, Victor; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vicente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Lopez, Jairo Alexis; Rodriguez Perez, Pablo; Rogozhnikov, Alexey; Roiser, Stefan; Romanovskiy, Vladimir; Romero Vidal, Antonio; Ronayne, John William; Rotondo, Marcello; Ruf, Thomas; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santimaria, Marco; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrina, Darya; Schael, Stefan; Schellenberg, Margarete; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmelzer, Timon; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schubert, Konstantin; Schubiger, Maxime; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Semennikov, Alexander; Sergi, Antonino; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Siddi, Benedetto Gianluca; Silva Coutinho, Rafael; Silva de Oliveira, Luiz Gustavo; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Eluned; Smith, Iwan Thomas; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Stefko, Pavol; Stefkova, Slavorima; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Syropoulos, Vasileios; Szczekowski, Marek; Szumlak, Tomasz; T'Jampens, Stephane; Tayduganov, Andrey; Tekampe, Tobias; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Tournefier, Edwige; Tourneur, Stephane; Trabelsi, Karim; Traill, Murdo; Tran, Minh Tâm; Tresch, Marco; Trisovic, Ana; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valat, Sebastien; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vecchi, Stefania; van Veghel, Maarten; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Venkateswaran, Aravindhan; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Volkov, Vladimir; Vollhardt, Achim; Voneki, Balazs; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; de Vries, Jacco; Vázquez Sierra, Carlos; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Weiden, Andreas; Whitehead, Mark; Wicht, Jean; Wilkinson, Guy; Wilkinson, Michael; Williams, Mark Richard James; Williams, Matthew; Williams, Mike; Williams, Timothy; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wraight, Kenneth; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xing, Zhou; Xu, Zhirui; Yang, Zhenwei; Yin, Hang; Yu, Jiesheng; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zarebski, Kristian Alexander; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Yanxi; Zhang, Yu; Zhelezov, Alexey; Zheng, Yangheng; Zhokhov, Anatoly; Zhukov, Valery; Zucchelli, Stefano

    2016-11-08

    A measurement of the differential branching fraction of the decay ${B^{0}\\rightarrow K^{\\ast}(892)^{0}\\mu^{+}\\mu^{-}}$ is presented together with a determination of the S-wave fraction of the $K^+\\pi^-$ system in the decay $B^{0}\\rightarrow K^{+}\\pi^{-}\\mu^{+}\\mu^{-}$. The analysis is based on $pp$-collision data corresponding to an integrated luminosity of 3\\,fb$^{-1}$ collected with the LHCb experiment. The measurements are made in bins of the invariant mass squared of the dimuon system, $q^2$. Precise theoretical predictions for the differential branching fraction of $B^{0}\\rightarrow K^{\\ast}(892)^{0}\\mu^{+}\\mu^{-}$ decays are available for the $q^2$ region $1.1

  16. IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

    International Nuclear Information System (INIS)

    Frajndlich, Roberto

    2009-01-01

    The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

  17. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

    1992-01-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

  18. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  19. Spent fuel management - two alternatives at the FiR 1 reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S.E.J.

    2001-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

  20. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

    1986-01-01

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

  1. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    International Nuclear Information System (INIS)

    Ansari, S.A.

    1990-01-01

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

  2. Properties of autoregressive model in reactor noise analysis, 1

    International Nuclear Information System (INIS)

    Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

    1987-01-01

    Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

  3. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1995-03-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

  4. Spent fuel management plans for the FiR 1 Reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S. E. J.

    2002-01-01

    The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

  5. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    International Nuclear Information System (INIS)

    Mesquita, Henrique F.A.; Ferreira, Andrea V.

    2015-01-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  6. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  7. RA reactor operation and maintenance in 1996, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1996-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  8. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  9. The AMPS 1.5 MW low-pressure compact reactor

    International Nuclear Information System (INIS)

    Hewitt, J.S.

    1987-01-01

    The 1.5-MWt reactor of the Autonomous Marine Power Source (AMPS) is designed to meet the unusual requirements of its first application. To provide for 100 kWe (net) on board self-sustaining manned submersible vehicles, the AMPS reactor must deliver safely, reliably and without direct operator surveillance, its thermal output to freon Rankine-cycle engines at thermodynamically useful temperatures. It must also conform to space and weight limits on the order of less than 50 cubic metres and 70 tonnes. The safety requirements are met by (i) limiting lifetime excess reactivity requirements by incorporation of burnable poison in the U-Zr-H fuel, (ii) maintaining nominal pressures in the light-water primary system at about 1 atmosphere, and (iii) maintaining a large volume of primary reserve coolant at temperature depressed relative to that of the circulating coolant. The latter averages 90 degrees celsius as it is pumped around loops that include the reactor core and the freon evaporators during normal operation. In the event of loss of pumped flow, the system defaults by intrinsic means to core cooling through natural convective exchange with the reserve coolant. In the post-shutdown situation, this passive cooling mode continues to operate regardless of vessel orientation and decay heat is safely dissipated to the sea. The design of the AMPS system, including the reactor, the freon engines, the control and monitoring system, the safety shut-down system and the power source container, are in advanced stages of design. (author)

  10. Flux distribution measurements in the Bruce A unit 1 reactor

    International Nuclear Information System (INIS)

    Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

    1977-07-01

    Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

  11. Prognostic value of pretreatment serum alanine aminotransferase/aspartate aminotransferase (ALT/AST) ratio and gamma glutamyltransferase (GGT) in patients with esophageal squamous cell carcinoma.

    Science.gov (United States)

    Huang, Hao; Wang, Xue-Ping; Li, Xiao-Hui; Chen, Hao; Zheng, Xin; Lin, Jian-Hua; Kang, Ting; Zhang, Lin; Chen, Pei-Song

    2017-08-14

    The levels of liver function tests (LFTs) are often used to assess liver injury and non-liver disease-related mortality. In our study, the relationship between pretreatment serum LFTs and overall survival (OS) was evaluated in esophageal squamous cell carcinoma (ESCC) patients. Our purpose was to investigate the prognostic value of the preoperative alanine aminotransferase/aspartate aminotransferase (ALT/AST) ratio and gamma glutamyltransferase (GGT) in ESCC patients. A retrospective study was performed in 447 patients with ESCC, and follow-up period was at least 60 months until death. The prognostic significance of serum LFTs were determined by univariate and multivariate Cox hazard models. LFTs including ALT, AST, LSR, GGT, TBA and LDH were analyzed. Serum LSR (HR: 0.592, 95% CI = 0.457-0.768, p < 0.001 and GGT (HR: 1.507, 95% CI = 1.163-1.953, p = 0.002) levels were indicated as significant predictors of OS. The 5-year OS among patients with higher LSR levels was longer compared with those patients with decreased LSR levels, not only in the whole cohort but also in the subgroups stratified by pathological stage (T1-T2 subgroup, T3-T4 subgroup, N0 subgroup and M0 subgroup). We also found that patients with a higher GGT might predict worse OS than patients with a normal GGT, not only in the whole cohort but also in the subgroups stratified by pathological stage (T3-T4 subgroup and N1-N2 subgroup). Both increased levels of LSR and decreased levels of GGT might predict shorter overall survival in ESCC patients. Our findings suggest that serum LSR and GGT levels could be used as a key predictor of survival in patients with ESCC.

  12. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2002-07-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  13. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    Science.gov (United States)

    El-Genk, Mohamed S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

  14. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    International Nuclear Information System (INIS)

    El-genk, M.S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

  15. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

  16. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

  17. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1); CAC-RA-1 1958-1998. Los primeros anios del CAC. Historia del primer reactor nuclear argentino (RA-1)

    Energy Technology Data Exchange (ETDEWEB)

    Forlerer, Elena; Palacios, Tulio A [comps.

    1998-07-01

    After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation.

  18. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  19. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  20. Use of self powered neutron detectors in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Galo Rocha, F. del.

    1989-01-01

    A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

  1. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    Energy Technology Data Exchange (ETDEWEB)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

    1999-08-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  2. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    International Nuclear Information System (INIS)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

    1999-01-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  3. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  4. Modifications done in the IPR-R1 reactor and their auxiliary systems

    International Nuclear Information System (INIS)

    Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

    1986-01-01

    The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

  5. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  6. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  7. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  8. Instrumentation renewal at the FIR 1 research reactor in Finland

    International Nuclear Information System (INIS)

    Bars, Bruno; Kall, Leif

    1982-01-01

    The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

  9. Dose measurements in controlled area of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, F.L.; Junior, F.M.

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers

  10. Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

    2003-06-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

  11. Equipment for neutron measurements at VR-1 Sparrow training reactor

    International Nuclear Information System (INIS)

    Kolros, Antonin; Huml, Ondrej; Kos, Josef

    2008-01-01

    Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

  12. Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

    International Nuclear Information System (INIS)

    Olsson, S.; Andermo, L.

    1980-01-01

    This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

  13. Measurement of the forward-backward asymmetry in $Z/\\gamma^{\\ast} \\rightarrow \\mu^{+}\\mu^{-}$ decays and determination of the effective weak mixing angle

    CERN Document Server

    Aaij, Roel; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreassi, Guido; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; d'Argent, Philippe; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Bel, Lennaert; Bellee, Violaine; Belloli, Nicoletta; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bertolin, Alessandro; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Billoir, Pierre; Bird, Thomas; Birnkraut, Alex; Bizzeti, Andrea; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Braun, Svende; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Buchanan, Emma; Bursche, Albert; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Capriotti, Lorenzo; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carniti, Paolo; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cavallero, Giovanni; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cogoni, Violetta; Cojocariu, Lucian; Collazuol, Gianmaria; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Crocombe, Andrew; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dall'Occo, Elena; Dalseno, Jeremy; David, Pieter; Davis, Adam; De Aguiar Francisco, Oscar; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Simone, Patrizia; Dean, Cameron Thomas; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Demmer, Moritz; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Dey, Biplab; Di Canto, Angelo; Di Ruscio, Francesco; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dossett, David; Dovbnya, Anatoliy; Dreimanis, Karlis; Dufour, Laurent; Dujany, Giulio; Dupertuis, Frederic; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferrari, Fabio; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fohl, Klaus; Fol, Philip; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garra Tico, Jordi; Garrido, Lluis; Gascon, David; Gaspar, Clara; Gauld, Rhorry; Gavardi, Laura; Gazzoni, Giulio; Gerick, David; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianì, Sebastiana; Gibson, Valerie; Girard, Olivier Göran; Giubega, Lavinia-Helena; Gligorov, V.V.; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graverini, Elena; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Gregson, Sam; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadavizadeh, Thomas; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Hennessy, Karol; Henrard, Pierre; Henry, Louis; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Humair, Thibaud; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kecke, Matthieu; Kelsey, Matthew; Kenyon, Ian; Kenzie, Matthew; Ketel, Tjeerd; Khairullin, Egor; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Kozeiha, Mohamad; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Krzemien, Wojciech; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kuonen, Axel Kevin; Kurek, Krzysztof; Kvaratskheliya, Tengiz; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Lemos Cid, Edgar; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Liu, Xuesong; Loh, David; Longstaff, Iain; Lopes, Jose; Lucchesi, Donatella; Lucio Martinez, Miriam; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Lusiani, Alberto; Machefert, Frederic; Maciuc, Florin; Maev, Oleg; Maguire, Kevin; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Manning, Peter Michael; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Marks, Jörg; Martellotti, Giuseppe; Martin, Morgan; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathad, Abhijit; Mathe, Zoltan; Matteuzzi, Clara; Mauri, Andrea; Maurin, Brice; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; Meadows, Brian; Meier, Frank; Meissner, Marco; Melnychuk, Dmytro; Merk, Marcel; Michielin, Emanuele; Milanes, Diego Alejandro; Minard, Marie-Noelle; Mitzel, Dominik Stefan; Molina Rodriguez, Josue; Monroy, Ignacio Alberto; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Dominik; Müller, Janine; Müller, Katharina; Müller, Vanessa; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nandi, Anita; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Osorio Rodrigues, Bruno; Otalora Goicochea, Juan Martin; Otto, Adam; Owen, Patrick; Oyanguren, Maria Aranzazu; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Pappenheimer, Cheryl; Parker, William; Parkes, Christopher; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pearce, Alex; Pellegrino, Antonio; Penso, Gianni; Pepe Altarelli, Monica; Perazzini, Stefano; Perret, Pascal; Pescatore, Luca; Petridis, Konstantinos; Petrolini, Alessandro; Petruzzo, Marco; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Piucci, Alessio; Playfer, Stephen; Plo Casasus, Maximo; Poikela, Tuomas; Polci, Francesco; Poluektov, Anton; Polyakov, Ivan; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Quagliani, Renato; Rachwal, Bartolomiej; Rademacker, Jonas; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Lopez, Jairo Alexis; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Ronayne, John William; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santimaria, Marco; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmelzer, Timon; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schubiger, Maxime; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Semennikov, Alexander; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Siddi, Benedetto Gianluca; Silva Coutinho, Rafael; Silva de Oliveira, Luiz Gustavo; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Edmund; Smith, Eluned; Smith, Iwan Thomas; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Stefkova, Slavorima; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szumlak, Tomasz; T'Jampens, Stephane; Tayduganov, Andrey; Tekampe, Tobias; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Todd, Jacob; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Trabelsi, Karim; Tran, Minh Tâm; Tresch, Marco; Trisovic, Ana; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Volkov, Vladimir; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Weiden, Andreas; Whitehead, Mark; Wilkinson, Guy; Wilkinson, Michael; Williams, Mark Richard James; Williams, Matthew; Williams, Mike; Williams, Timothy; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xu, Zhirui; Yang, Zhenwei; Yu, Jiesheng; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang; Zucchelli, Stefano

    2015-11-27

    The forward-backward charge asymmetry for the process $q\\bar{q} \\rightarrow Z/\\gamma^{\\ast} \\rightarrow \\mu^{+}\\mu^{-}$ is measured as a function of the invariant mass of the dimuon system. Measurements are performed using proton proton collision data collected with the LHCb detector at $\\sqrt {s} = 7$ and 8 TeV, corresponding to integrated luminosities of 1 fb$^{-1}$ and 2 fb$^{-1}$ respectively. Within the Standard Model the results constrain the effective electroweak mixing angle to be $\\text{sin} ^2\\theta ^\\text{eff} _W = 0.23142 \\pm 0.00073 \\pm 0.00052 \\pm 0.00056$, where the first uncertainty is statistical, the second systematic and the third theoretical. This result is in agreement with the current world average, and is one of the most precise determinations at hadron colliders to date.

  14. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1986-01-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  15. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  16. Integrated Variable-Fidelity Tool Set for Modeling and Simulation of Aeroservothermoelasticity-Propulsion (ASTE-P) Effects for Aerospace Vehicles Ranging From Subsonic to Hypersonic Flight, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed research program aims at developing a variable-fidelity software tool set for aeroservothermoelastic-propulsive (ASTE-P) modeling that can be routinely...

  17. Integrated Variable-Fidelity Tool Set For Modeling and Simulation of Aeroservothermoelasticity -Propulsion (ASTE-P) Effects For Aerospace Vehicles Ranging From Subsonic to Hypersonic Flight, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed research program aims at developing a variable-fidelity software tool set for aeroservothermoelastic-propulsive (ASTE-P) modeling that can be routinely...

  18. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1988-01-01

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  19. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

    2008-01-01

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  20. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  1. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    2002-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

  2. Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

    International Nuclear Information System (INIS)

    Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

    2013-01-01

    The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

  3. Expanding the storage capability at ET-RR-1 research reactor at Inshass

    International Nuclear Information System (INIS)

    Sultan, Mariy M.; Khattab, M.

    1999-01-01

    Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

  4. Licensing of the first reload of Angra-1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

  5. Evaluation of the trial design studies for an advanced marine reactor, (1)

    International Nuclear Information System (INIS)

    1988-03-01

    The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

  6. Transformation of 1,1,1-trichloroethane in an anaerobic packed-bed reactor at various concentrations of 1,1,1-trichloroethane, acetate and sulfate

    NARCIS (Netherlands)

    deBest, JH; Jongema, H; Weijling, A; Doddema, HJ; Janssen, DB; Harder, W

    Biotransformation of 1,1,1-trichloroethane (CH3CCl3) was observed in an anaerobic packed-bed reactor under conditions of both sulfate reduction and methanogenesis. Acetate (1 mM) served as an electron donor. CH3CCl3 was completely converted up to the highest investigated concentration of 10 mu M.

  7. Expressão do Mg+2, CK, AST e LDH em equinos finalistas de provas de enduro

    Directory of Open Access Journals (Sweden)

    Juliana V.F. Sales

    2013-01-01

    Full Text Available Nos últimos anos, o equino atleta vem sendo cada vez mais requerido. Dessa forma, as exigências por alto desempenho têm fomentado o interesse pelo estudo das afecções relacionadas com a fisiopatologia de diversas enfermidades dos equinos. A relação entre o íon magnésio e o exercício físico tem recebido atenção significativa visto que este íon está intimamente relacionado ao tecido muscular estriado esquelético. Além disso, dentre as principais estratégias para a detecção e acompanhamento clínico de lesões musculares, destacam-se a avaliação das atividades das enzimas creatino quinase (CK, lactato desidrogenase (LDH e aspartato aminotransferase (AST. A busca pelo estabelecimento de parâmetros que se relacionam entre si é um fator determinante na compreensão de alterações fisiológicas encontradas diante do esforço em equinos atletas. Desta forma, o presente trabalho teve como objetivo determinar como as concentrações sanguíneas do íon magnésio e as atividades enzimáticas das enzimas CK, LDH e AST comportaram-se em equinos Puro Sangue Árabe finalistas de provas de enduro de 90km e relacionar as possíveis alterações com o tipo de esforço físico desempenhado pelos animais. Foram avaliadas a atividade enzimática das enzimas CK, LDH, AST e a concentração do íon magnésio no exercício em relação ao repouso de 14 equinos clinicamente hígidos da raça Puro Sangue Árabe, sendo 9 machos e 5 fêmeas, com idades variando entre 6 a 12 anos, submetidos a treinamento para enduro e participantes de provas de 90 km. Pode-se observar que as variáveis acima mencionadas sofreram aumento com diferença estatística em relação ao repouso. O exercício físico de enduro determinou a ocorrência de alterações nas atividades enzimáticas das enzimas CK (p≤0,001, LDH (p=0,0001, AST (p=0,0007 e na concentração do íon magnésio (p=0,0004, no exercício em relação ao repouso (p≤0,05. Fato que determinou altera

  8. Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

    International Nuclear Information System (INIS)

    Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

    2009-01-01

    The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

  9. Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

    2014-08-15

    The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

  10. Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1985-01-01

    This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

  11. Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

    International Nuclear Information System (INIS)

    Andermo, L.; Sundman, B.

    1974-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

  12. New digital control system for the operation of the Colombian research reactor IAN-R1

    International Nuclear Information System (INIS)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

    2015-09-01

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  13. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  14. Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

    International Nuclear Information System (INIS)

    Andermo, L.

    1977-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

  15. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    Energy Technology Data Exchange (ETDEWEB)

    El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

  16. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    International Nuclear Information System (INIS)

    El-Messeiry, A.M.

    1996-01-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

  17. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

  18. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2007-12-15

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  19. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  20. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-01-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  1. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  2. Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

    Science.gov (United States)

    Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

    2017-11-01

    The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

  3. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C D [comp.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

  4. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  5. Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

  6. Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

    International Nuclear Information System (INIS)

    Tikhomirov, L.N.

    1995-01-01

    The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

  7. RA reactor operation and maintenance in 1994, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1994-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  8. Investigation of the polyvinyl alcohol stabilization mechanism and adsorption properties on the surface of ternary mixed nanooxide AST 50 (Al{sub 2}O{sub 3}–SiO{sub 2}–TiO{sub 2})

    Energy Technology Data Exchange (ETDEWEB)

    Wiśniewska, Małgorzata; Ostolska, Iwona, E-mail: i-ostolska@wp.pl; Szewczuk-Karpisz, Katarzyna; Chibowski, Stanisław [Maria Curie-Sklodowska University, Department of Radiochemistry and Colloids Chemistry, Faculty of Chemistry (Poland); Terpiłowski, Konrad [Maria Curie-Sklodowska University, Department of Physical Chemistry – Interfacial Phenomena, Faculty of Chemistry (Poland); Gun’ko, Vladimir Moiseevich; Zarko, Vladimir Iljich [National Academy of Sciences in Ukraine, Institute of Surface Chemistry (Ukraine)

    2015-01-15

    A new adsorbent consisting of fumed, mixed alumina, silica, and titania in various proportions (AST 50) was investigated. The studied material was prepared by chemical vapor deposition method. The diameter of AST 50 primary particles was equal to about 51 nm which denotes that it can be classified as a nanomaterial. In the presented paper, the adsorption properties of polyvinyl alcohol on the ternary oxide were investigated. The polymer macromolecules were characterized by two different molecular weights and degree of hydrolysis. The polymer adsorption reaches the maximum at pH 3 and decreases with the solution pH rise. The reduction of the adsorbed PVA macromolecules is related to the electrostatic repulsion forces occurring in the studied system. The AST 50 point of zero charge (pH{sub pzc}) obtained from the potentiometric titration is equal to 4.7. Due to the nonionic character of the analyzed macromolecular compound, the polymer attendance has an insignificant effect on the AST 50 surface charge density. In the case of the adsorbent particles zeta potential, the obtained dependencies are different in the absence and presence of PVA. The shift of the slipping plane and displacement of the counter-ions from Stern layer by the adsorbed polymer chains have the greatest effect on the ζ potential value. The stability measurements indicate that the AST 50 suspensions in the presence of the background electrolyte at pH 3 and 6 are unstable. In turn, in an alkaline medium the mixed oxide suspensions exhibit the highest durability, which is a result of a large number of the negative charges on the AST 50 surface. The addition of PVA 100 significantly improves the suspension stability at pH 3 and 6; at higher pH value, the polymer presence does not influence the system durability. It is related to the steric and electrosteric stabilization of the colloidal particles by the adsorbed polyvinyl alcohol macromolecules.

  9. PARÁMETROS BIOQUÍMICOS ENZIMÁTICOS (ALT, AST, ALP, Γ-GT, LDH EN NIÑOS CON LEUCEMIA LINFOBLÁSTICA AGUDA ANTES DEL TRATAMIENTO ANTINEOPLÁSICO

    Directory of Open Access Journals (Sweden)

    Jeél Moya S

    2015-12-01

    Full Text Available Objective: To determine the enzymatic biochemical parameters (glutamic pyruvic transaminase (ALT, glutamic oxaloacetic transaminase (AST, alkaline phosphatase (ALP, gamma glutamyltransferase (γ-GT, and lactate dehydrogenase (LDH in children with acute lymphoblastic leukemia (ALL before cancer treatment. Material and Methods: A prospective experimental, observational, cross-sectional study was conducted in 30 children between 2 and 15 years old, from several Neoplastic Centers in Lima. Blood collection was performed in BD red cap Vacutainer tubes, processed in the semi-automated analyzer BIOTEC® EMP-168, with Wiener Lab Group enzyme reagents under the modified method Szaaz and UV-Optimized by IFCC, SSCC and SFBC. Finally, coding and tabulation was performed. Results: 60% were boys and 46.7% are between the ages of 2-6 years. Serum levels of AST were increased by 33.3% in boys and 50% in girls. Serum ALT values were increased in 33.3% of boys and 41.7% of girls; only 25% of girls showed increased levels of γ-GT values; ALP was increased in 44.4% of boys and 66.7% of girls. Moreover LDH levels were increased in 55.6% of boys and 41.7% of girls. Conclusions: The enzymatic tests LDH, AST, ALT and ALP are increased in children with ALL compared to normal values due to tumor lysis syndrome characterized by electrolyte abnormalities, and as a result of the massive destruction of tumor cells and rapid release of large amounts of intracellular elements.

  10. Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Manas Diaz, L; Montes Ponce de leon, J.

    1960-07-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

  11. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  12. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  13. Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

    International Nuclear Information System (INIS)

    1986-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

  14. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

  15. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  16. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Shal'nov, A.V.

    1995-01-01

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  17. Experiment on continuous operation of the Brazilian IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Freitas Pintaud, M. de

    1994-01-01

    In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

  18. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    Coura, J.G.

    1986-01-01

    This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

  19. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  20. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    International Nuclear Information System (INIS)

    Novelli, A.

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

  1. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    International Nuclear Information System (INIS)

    Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

    2008-01-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  2. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

  3. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2000-01-01

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

  4. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F

    1973-07-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  5. Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

    International Nuclear Information System (INIS)

    1987-08-01

    Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

  6. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  7. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  8. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  9. Modelling of the RA-1 reactor using a Monte Carlo code; Modelado del reactor RA-1 utilizando un codigo Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Quinteiro, Guillermo F; Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Reactores y Centrales Nucleares

    2000-07-01

    It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

  10. Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

  11. Applied research into direct numerical control of A-1 reactor temperature

    International Nuclear Information System (INIS)

    Karpeta, C.; Volf, K.

    1974-01-01

    Partial results of research efforts aimed at applying modern control theory in the control of the reactor of the A-1 nuclear power station are presented. A mathematical model of the process dynamics was developed. Some parameters of the model were determined using the results of an experimentally performed reactor scram. The optimal stochastic discrete regulator was determined and closed-loop transients were studied. The possibilities of implementing control routines were investigated using the RPP-16 computer. (author)

  12. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-01-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  13. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  14. Research reactor core conversion guidebook. V.1: Summary

    International Nuclear Information System (INIS)

    1992-04-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

  15. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits

  16. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  17. NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

    International Nuclear Information System (INIS)

    Paiva, Rosemeire P.; Salvetti, Tereza C.

    2005-01-01

    Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

  18. Life extension of the St. Lucie unit 1 reactor vessel

    International Nuclear Information System (INIS)

    Rowan, G.A.; Sun, J.B.; Mott, S.L.

    1991-01-01

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

  19. Reactor calculations in aid of isotope production at SAFARI-1

    International Nuclear Information System (INIS)

    Ball, G.

    2003-01-01

    Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

  20. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  1. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    Chambadal, P.; Pascal, M.

    1955-01-01

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr

  2. Development of a training simulator to operators of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Carvalho, Ricardo Pinto de

    2006-01-01

    This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

  3. An overview of the RECH-1 reactor conversion

    International Nuclear Information System (INIS)

    Klein, J.; Medel, J.; Daie, J.; Torres, H.

    2000-01-01

    The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

  4. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  5. Verification of the linearity of the IPR-R1 TRIGA reactor power channels

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

    2013-01-01

    The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

  6. Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

    2011-01-01

    JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

  7. Study of $(W/Z)H$ production and Higgs boson couplings using $H \\rightarrow WW^{\\ast}$ decays with the ATLAS detector

    CERN Document Server

    Aad, Georges; Abdallah, Jalal; Abdinov, Ovsat; Aben, Rosemarie; Abolins, Maris; AbouZeid, Ossama; Abramowicz, Halina; Abreu, Henso; Abreu, Ricardo; Abulaiti, Yiming; Acharya, Bobby Samir; Adamczyk, Leszek; Adams, David; Adelman, Jahred; Adomeit, Stefanie; Adye, Tim; Affolder, Tony; Agatonovic-Jovin, Tatjana; Aguilar-Saavedra, Juan Antonio; Ahlen, Steven; Ahmadov, Faig; Aielli, Giulio; Akerstedt, Henrik; Åkesson, Torsten Paul Ake; Akimoto, Ginga; Akimov, Andrei; Alberghi, Gian Luigi; Albert, Justin; Albrand, Solveig; Alconada Verzini, Maria Josefina; Aleksa, Martin; Aleksandrov, Igor; Alexa, Calin; Alexander, Gideon; Alexopoulos, Theodoros; Alhroob, Muhammad; Alimonti, Gianluca; Alio, Lion; Alison, John; Alkire, Steven Patrick; Allbrooke, Benedict; Allport, Phillip; Aloisio, Alberto; Alonso, Alejandro; Alonso, Francisco; Alpigiani, Cristiano; Altheimer, Andrew David; Alvarez Gonzalez, Barbara; Άlvarez Piqueras, Damián; Alviggi, Mariagrazia; Amadio, Brian Thomas; Amako, Katsuya; Amaral Coutinho, Yara; Amelung, Christoph; Amidei, Dante; Amor Dos Santos, Susana Patricia; Amorim, Antonio; Amoroso, Simone; Amram, Nir; Amundsen, Glenn; Anastopoulos, Christos; Ancu, Lucian Stefan; Andari, Nansi; Andeen, Timothy; Anders, Christoph Falk; Anders, Gabriel; Anders, John Kenneth; Anderson, Kelby; Andreazza, Attilio; Andrei, George Victor; Angelidakis, Stylianos; Angelozzi, Ivan; Anger, Philipp; Angerami, Aaron; Anghinolfi, Francis; Anisenkov, Alexey; Anjos, Nuno; Annovi, Alberto; Antonelli, Mario; Antonov, Alexey; Antos, Jaroslav; Anulli, Fabio; Aoki, Masato; Aperio Bella, Ludovica; Arabidze, Giorgi; Arai, Yasuo; Araque, Juan Pedro; Arce, Ayana; Arduh, Francisco Anuar; Arguin, Jean-Francois; Argyropoulos, Spyridon; Arik, Metin; Armbruster, Aaron James; Arnaez, Olivier; Arnal, Vanessa; Arnold, Hannah; Arratia, Miguel; Arslan, Ozan; Artamonov, Andrei; Artoni, Giacomo; Asai, Shoji; Asbah, Nedaa; Ashkenazi, Adi; Åsman, Barbro; Asquith, Lily; Assamagan, Ketevi; Astalos, Robert; Atkinson, Markus; Atlay, Naim Bora; Auerbach, Benjamin; Augsten, Kamil; Aurousseau, Mathieu; Avolio, Giuseppe; Axen, Bradley; Ayoub, Mohamad Kassem; Azuelos, Georges; Baak, Max; Baas, Alessandra; Bacci, Cesare; Bachacou, Henri; Bachas, Konstantinos; Backes, Moritz; Backhaus, Malte; Bagiacchi, Paolo; Bagnaia, Paolo; Bai, Yu; Bain, Travis; Baines, John; Baker, Oliver Keith; Balek, Petr; Balestri, Thomas; Balli, Fabrice; Banas, Elzbieta; Banerjee, Swagato; Bannoura, Arwa A E; Bansil, Hardeep Singh; Barak, Liron; Barberio, Elisabetta Luigia; Barberis, Dario; Barbero, Marlon; Barillari, Teresa; Barisonzi, Marcello; Barklow, Timothy; Barlow, Nick; Barnes, Sarah Louise; Barnett, Bruce; Barnett, Michael; Barnovska, Zuzana; Baroncelli, Antonio; Barone, Gaetano; Barr, Alan; Barreiro, Fernando; Barreiro Guimarães da Costa, João; Bartoldus, Rainer; Barton, Adam Edward; Bartos, Pavol; Basalaev, Artem; Bassalat, Ahmed; Basye, Austin; Bates, Richard; Batista, Santiago Juan; Batley, Richard; Battaglia, Marco; Bauce, Matteo; Bauer, Florian; Bawa, Harinder Singh; Beacham, James Baker; Beattie, Michael David; Beau, Tristan; Beauchemin, Pierre-Hugues; Beccherle, Roberto; Bechtle, Philip; Beck, Hans Peter; Becker, Anne Kathrin; Becker, Maurice; Becker, Sebastian; Beckingham, Matthew; Becot, Cyril; Beddall, Andrew; Beddall, Ayda; Bednyakov, Vadim; Bee, Christopher; Beemster, Lars; Beermann, Thomas; Begel, Michael; Behr, Janna Katharina; Belanger-Champagne, Camille; Bell, William; Bella, Gideon; Bellagamba, Lorenzo; Bellerive, Alain; Bellomo, Massimiliano; Belotskiy, Konstantin; Beltramello, Olga; Benary, Odette; Benchekroun, Driss; Bender, Michael; Bendtz, Katarina; Benekos, Nektarios; Benhammou, Yan; Benhar Noccioli, Eleonora; Benitez Garcia, Jorge-Armando; Benjamin, Douglas; Bensinger, James; Bentvelsen, Stan; Beresford, Lydia; Beretta, Matteo; Berge, David; Bergeaas Kuutmann, Elin; Berger, Nicolas; Berghaus, Frank; Beringer, Jürg; Bernard, Clare; Bernard, Nathan Rogers; Bernius, Catrin; Bernlochner, Florian Urs; Berry, Tracey; Berta, Peter; Bertella, Claudia; Bertoli, Gabriele; Bertolucci, Federico; Bertsche, Carolyn; Bertsche, David; Besana, Maria Ilaria; Besjes, Geert-Jan; Bessidskaia Bylund, Olga; Bessner, Martin Florian; Besson, Nathalie; Betancourt, Christopher; Bethke, Siegfried; Bevan, Adrian John; Bhimji, Wahid; Bianchi, Riccardo-Maria; Bianchini, Louis; Bianco, Michele; Biebel, Otmar; Bieniek, Stephen Paul; Biglietti, Michela; Bilbao De Mendizabal, Javier; Bilokon, Halina; Bindi, Marcello; Binet, Sebastien; Bingul, Ahmet; Bini, Cesare; Black, Curtis; Black, James; Black, Kevin; Blackburn, Daniel; Blair, Robert; Blanchard, Jean-Baptiste; Blanco, Jacobo Ezequiel; Blazek, Tomas; Bloch, Ingo; Blocker, Craig; Blum, Walter; Blumenschein, Ulrike; Bobbink, Gerjan; Bobrovnikov, Victor; Bocchetta, Simona Serena; Bocci, Andrea; Bock, Christopher; Boehler, Michael; Bogaerts, Joannes Andreas; Bogdanchikov, Alexander; Bohm, Christian; Boisvert, Veronique; Bold, Tomasz; Boldea, Venera; Boldyrev, Alexey; Bomben, Marco; Bona, Marcella; Boonekamp, Maarten; Borisov, Anatoly; Borissov, Guennadi; Borroni, Sara; Bortfeldt, Jonathan; Bortolotto, Valerio; Bos, Kors; Boscherini, Davide; Bosman, Martine; Boudreau, Joseph; Bouffard, Julian; Bouhova-Thacker, Evelina Vassileva; Boumediene, Djamel Eddine; Bourdarios, Claire; Bousson, Nicolas; Boveia, Antonio; Boyd, James; Boyko, Igor; Bozic, Ivan; Bracinik, Juraj; Brandt, Andrew; Brandt, Gerhard; Brandt, Oleg; Bratzler, Uwe; Brau, Benjamin; Brau, James; Braun, Helmut; Brazzale, Simone Federico; Brendlinger, Kurt; Brennan, Amelia Jean; Brenner, Lydia; Brenner, Richard; Bressler, Shikma; Bristow, Kieran; Bristow, Timothy Michael; Britton, Dave; Britzger, Daniel; Brochu, Frederic; Brock, Ian; Brock, Raymond; Bronner, Johanna; Brooijmans, Gustaaf; Brooks, Timothy; Brooks, William; Brosamer, Jacquelyn; Brost, Elizabeth; Brown, Jonathan; Bruckman de Renstrom, Pawel; Bruncko, Dusan; Bruneliere, Renaud; Bruni, Alessia; Bruni, Graziano; Bruschi, Marco; Bryngemark, Lene; Buanes, Trygve; Buat, Quentin; Buchholz, Peter; Buckley, Andrew; Buda, Stelian Ioan; Budagov, Ioulian; Buehrer, Felix; Bugge, Lars; Bugge, Magnar Kopangen; Bulekov, Oleg; Bullock, Daniel; Burckhart, Helfried; Burdin, Sergey; Burghgrave, Blake; Burke, Stephen; Burmeister, Ingo; Busato, Emmanuel; Büscher, Daniel; Büscher, Volker; Bussey, Peter; Butler, John; Butt, Aatif Imtiaz; Buttar, Craig; Butterworth, Jonathan; Butti, Pierfrancesco; Buttinger, William; Buzatu, Adrian; Buzykaev, Aleksey; Cabrera Urbán, Susana; Caforio, Davide; Cairo, Valentina; Cakir, Orhan; Calafiura, Paolo; Calandri, Alessandro; Calderini, Giovanni; Calfayan, Philippe; Caloba, Luiz; Calvet, David; Calvet, Samuel; Camacho Toro, Reina; Camarda, Stefano; Camarri, Paolo; Cameron, David; Caminada, Lea Michaela; Caminal Armadans, Roger; Campana, Simone; Campanelli, Mario; Campoverde, Angel; Canale, Vincenzo; Canepa, Anadi; Cano Bret, Marc; Cantero, Josu; Cantrill, Robert; Cao, Tingting; Capeans Garrido, Maria Del Mar; Caprini, Irinel; Caprini, Mihai; Capua, Marcella; Caputo, Regina; Cardarelli, Roberto; Carli, Tancredi; Carlino, Gianpaolo; Carminati, Leonardo; Caron, Sascha; Carquin, Edson; Carrillo-Montoya, German D; Carter, Janet; Carvalho, João; Casadei, Diego; Casado, Maria Pilar; Casolino, Mirkoantonio; Castaneda-Miranda, Elizabeth; Castelli, Angelantonio; Castillo Gimenez, Victoria; Castro, Nuno Filipe; Catastini, Pierluigi; Catinaccio, Andrea; Catmore, James; Cattai, Ariella; Caudron, Julien; Cavaliere, Viviana; Cavalli, Donatella; Cavalli-Sforza, Matteo; Cavasinni, Vincenzo; Ceradini, Filippo; Cerio, Benjamin; Cerny, Karel; Santiago Cerqueira, Augusto; Cerri, Alessandro; Cerrito, Lucio; Cerutti, Fabio; Cerv, Matevz; Cervelli, Alberto; Cetin, Serkant Ali; Chafaq, Aziz; Chakraborty, Dhiman; Chalupkova, Ina; Chang, Philip; Chapleau, Bertrand; Chapman, John Derek; Charlton, Dave; Chau, Chav Chhiv; Chavez Barajas, Carlos Alberto; Cheatham, Susan; Chegwidden, Andrew; Chekanov, Sergei; Chekulaev, Sergey; Chelkov, Gueorgui; Chelstowska, Magda Anna; Chen, Chunhui; Chen, Hucheng; Chen, Karen; Chen, Liming; Chen, Shenjian; Chen, Xin; Chen, Ye; Cheng, Hok Chuen; Cheng, Yangyang; Cheplakov, Alexander; Cheremushkina, Evgenia; Cherkaoui El Moursli, Rajaa; Chernyatin, Valeriy; Cheu, Elliott; Chevalier, Laurent; Chiarella, Vitaliano; Childers, John Taylor; Chiodini, Gabriele; Chisholm, Andrew; Chislett, Rebecca Thalatta; Chitan, Adrian; Chizhov, Mihail; Choi, Kyungeon; Chouridou, Sofia; Chow, Bonnie Kar Bo; Christodoulou, Valentinos; Chromek-Burckhart, Doris; Chu, Ming-Lee; Chudoba, Jiri; Chuinard, Annabelle Julia; Chwastowski, Janusz; Chytka, Ladislav; Ciapetti, Guido; Ciftci, Abbas Kenan; Cinca, Diane; Cindro, Vladimir; Cioara, Irina Antonela; Ciocio, Alessandra; Citron, Zvi Hirsh; Ciubancan, Mihai; Clark, Allan G; Clark, Brian Lee; Clark, Philip James; Clarke, Robert; Cleland, Bill; Clement, Christophe; Coadou, Yann; Cobal, Marina; Coccaro, Andrea; Cochran, James H; Coffey, Laurel; Cogan, Joshua Godfrey; Cole, Brian; Cole, Stephen; Colijn, Auke-Pieter; Collot, Johann; Colombo, Tommaso; Compostella, Gabriele; Conde Muiño, Patricia; Coniavitis, Elias; Connell, Simon Henry; Connelly, Ian; Consonni, Sofia Maria; Consorti, Valerio; Constantinescu, Serban; Conta, Claudio; Conti, Geraldine; Conventi, Francesco; Cooke, Mark; Cooper, Ben; Cooper-Sarkar, Amanda; Cornelissen, Thijs; Corradi, Massimo; Corriveau, Francois; Corso-Radu, Alina; Cortes-Gonzalez, Arely; Cortiana, Giorgio; Costa, Giuseppe; Costa, María José; Costanzo, Davide; Côté, David; Cottin, Giovanna; Cowan, Glen; Cox, Brian; Cranmer, Kyle; Cree, Graham; Crépé-Renaudin, Sabine; Crescioli, Francesco; Cribbs, Wayne Allen; Crispin Ortuzar, Mireia; Cristinziani, Markus; Croft, Vince; Crosetti, Giovanni; Cuhadar Donszelmann, Tulay; Cummings, Jane; Curatolo, Maria; Cuthbert, Cameron; Czirr, Hendrik; Czodrowski, Patrick; D'Auria, Saverio; D'Onofrio, Monica; Da Cunha Sargedas De Sousa, Mario Jose; Da Via, Cinzia; Dabrowski, Wladyslaw; Dafinca, Alexandru; Dai, Tiesheng; Dale, Orjan; Dallaire, Frederick; Dallapiccola, Carlo; Dam, Mogens; Dandoy, Jeffrey Rogers; Dang, Nguyen Phuong; Daniells, Andrew Christopher; Danninger, Matthias; Dano Hoffmann, Maria; Dao, Valerio; Darbo, Giovanni; Darmora, Smita; Dassoulas, James; Dattagupta, Aparajita; Davey, Will; David, Claire; Davidek, Tomas; Davies, Eleanor; Davies, Merlin; Davison, Peter; Davygora, Yuriy; Dawe, Edmund; Dawson, Ian; Daya-Ishmukhametova, Rozmin; De, Kaushik; de Asmundis, Riccardo; De Castro, Stefano; De Cecco, Sandro; De Groot, Nicolo; de Jong, Paul; De la Torre, Hector; De Lorenzi, Francesco; De Nooij, Lucie; De Pedis, Daniele; De Salvo, Alessandro; De Sanctis, Umberto; De Santo, Antonella; De Vivie De Regie, Jean-Baptiste; Dearnaley, William James; Debbe, Ramiro; Debenedetti, Chiara; Dedovich, Dmitri; Deigaard, Ingrid; Del Peso, Jose; Del Prete, Tarcisio; Delgove, David; Deliot, Frederic; Delitzsch, Chris Malena; Deliyergiyev, Maksym; Dell'Acqua, Andrea; Dell'Asta, Lidia; Dell'Orso, Mauro; Della Pietra, Massimo; della Volpe, Domenico; Delmastro, Marco; Delsart, Pierre-Antoine; Deluca, Carolina; DeMarco, David; Demers, Sarah; Demichev, Mikhail; Demilly, Aurelien; Denisov, Sergey; Derendarz, Dominik; Derkaoui, Jamal Eddine; Derue, Frederic; Dervan, Paul; Desch, Klaus Kurt; Deterre, Cecile; Deviveiros, Pier-Olivier; Dewhurst, Alastair; Dhaliwal, Saminder; Di Ciaccio, Anna; Di Ciaccio, Lucia; Di Domenico, Antonio; Di Donato, Camilla; Di Girolamo, Alessandro; Di Girolamo, Beniamino; Di Mattia, Alessandro; Di Micco, Biagio; Di Nardo, Roberto; Di Simone, Andrea; Di Sipio, Riccardo; Di Valentino, David; Diaconu, Cristinel; Diamond, Miriam; Dias, Flavia; Diaz, Marco Aurelio; Diehl, Edward; Dietrich, Janet; Diglio, Sara; Dimitrievska, Aleksandra; Dingfelder, Jochen; Dita, Petre; Dita, Sanda; Dittus, Fridolin; Djama, Fares; Djobava, Tamar; Djuvsland, Julia Isabell; Barros do Vale, Maria Aline; Dobos, Daniel; Dobre, Monica; Doglioni, Caterina; Dohmae, Takeshi; Dolejsi, Jiri; Dolezal, Zdenek; Dolgoshein, Boris; Donadelli, Marisilvia; Donati, Simone; Dondero, Paolo; Donini, Julien; Dopke, Jens; Doria, Alessandra; Dova, Maria-Teresa; Doyle, Tony; Drechsler, Eric; Dris, Manolis; Dubreuil, Emmanuelle; Duchovni, Ehud; Duckeck, Guenter; Ducu, Otilia Anamaria; Duda, Dominik; Dudarev, Alexey; Duflot, Laurent; Duguid, Liam; Dührssen, Michael; Dunford, Monica; Duran Yildiz, Hatice; Düren, Michael; Durglishvili, Archil; Duschinger, Dirk; Dyndal, Mateusz; Eckardt, Christoph; Ecker, Katharina Maria; Edgar, Ryan Christopher; Edson, William; Edwards, Nicholas Charles; Ehrenfeld, Wolfgang; Eifert, Till; Eigen, Gerald; Einsweiler, Kevin; Ekelof, Tord; El Kacimi, Mohamed; Ellert, Mattias; Elles, Sabine; Ellinghaus, Frank; Elliot, Alison; Ellis, Nicolas; Elmsheuser, Johannes; Elsing, Markus; Emeliyanov, Dmitry; Enari, Yuji; Endner, Oliver Chris; Endo, Masaki; Erdmann, Johannes; Ereditato, Antonio; Ernis, Gunar; Ernst, Jesse; Ernst, Michael; Errede, Steven; Ertel, Eugen; Escalier, Marc; Esch, Hendrik; Escobar, Carlos; Esposito, Bellisario; Etienvre, Anne-Isabelle; Etzion, Erez; Evans, Hal; Ezhilov, Alexey; Fabbri, Laura; Facini, Gabriel; Fakhrutdinov, Rinat; Falciano, Speranza; Falla, Rebecca Jane; Faltova, Jana; Fang, Yaquan; Fanti, Marcello; Farbin, Amir; Farilla, Addolorata; Farooque, Trisha; Farrell, Steven; Farrington, Sinead; Farthouat, Philippe; Fassi, Farida; Fassnacht, Patrick; Fassouliotis, Dimitrios; Faucci Giannelli, Michele; Favareto, Andrea; Fayard, Louis; Federic, Pavol; Fedin, Oleg; Fedorko, Wojciech; Feigl, Simon; Feligioni, Lorenzo; Feng, Cunfeng; Feng, Eric; Feng, Haolu; Fenyuk, Alexander; Fernandez Martinez, Patricia; Fernandez Perez, Sonia; Ferrando, James; Ferrari, Arnaud; Ferrari, Pamela; Ferrari, Roberto; Ferreira de Lima, Danilo Enoque; Ferrer, Antonio; Ferrere, Didier; Ferretti, Claudio; Ferretto Parodi, Andrea; Fiascaris, Maria; Fiedler, Frank; Filipčič, Andrej; Filipuzzi, Marco; Filthaut, Frank; Fincke-Keeler, Margret; Finelli, Kevin Daniel; Fiolhais, Miguel; Fiorini, Luca; Firan, Ana; Fischer, Adam; Fischer, Cora; Fischer, Julia; Fisher, Wade Cameron; Fitzgerald, Eric Andrew; Flechl, Martin; Fleck, Ivor; Fleischmann, Philipp; Fleischmann, Sebastian; Fletcher, Gareth Thomas; Fletcher, Gregory; Flick, Tobias; Floderus, Anders; Flores Castillo, Luis; Flowerdew, Michael; Formica, Andrea; Forti, Alessandra; Fournier, Daniel; Fox, Harald; Fracchia, Silvia; Francavilla, Paolo; Franchini, Matteo; Francis, David; Franconi, Laura; Franklin, Melissa; Fraternali, Marco; Freeborn, David; French, Sky; Friedrich, Felix; Froidevaux, Daniel; Frost, James; Fukunaga, Chikara; Fullana Torregrosa, Esteban; Fulsom, Bryan Gregory; Fuster, Juan; Gabaldon, Carolina; Gabizon, Ofir; Gabrielli, Alessandro; Gabrielli, Andrea; Gadatsch, Stefan; Gadomski, Szymon; Gagliardi, Guido; Gagnon, Pauline; Galea, Cristina; Galhardo, Bruno; Gallas, Elizabeth; Gallop, Bruce; Gallus, Petr; Galster, Gorm Aske Gram Krohn; Gan, KK; Gao, Jun; Gao, Yanyan; Gao, Yongsheng; Garay Walls, Francisca; Garberson, Ford; García, Carmen; García Navarro, José Enrique; Garcia-Sciveres, Maurice; Gardner, Robert; Garelli, Nicoletta; Garonne, Vincent; Gatti, Claudio; Gaudiello, Andrea; Gaudio, Gabriella; Gaur, Bakul; Gauthier, Lea; Gauzzi, Paolo; Gavrilenko, Igor; Gay, Colin; Gaycken, Goetz; Gazis, Evangelos; Ge, Peng; Gecse, Zoltan; Gee, Norman; Geerts, Daniël Alphonsus Adrianus; Geich-Gimbel, Christoph; Geisler, Manuel Patrice; Gemme, Claudia; Genest, Marie-Hélène; Gentile, Simonetta; George, Matthias; George, Simon; Gerbaudo, Davide; Gershon, Avi; Ghazlane, Hamid; Giacobbe, Benedetto; Giagu, Stefano; Giangiobbe, Vincent; Giannetti, Paola; Gibbard, Bruce; Gibson, Stephen; Gilchriese, Murdock; Gillam, Thomas; Gillberg, Dag; Gilles, Geoffrey; Gingrich, Douglas; Giokaris, Nikos; Giordani, MarioPaolo; Giorgi, Filippo Maria; Giorgi, Francesco Michelangelo; Giraud, Pierre-Francois; Giromini, Paolo; Giugni, Danilo; Giuliani, Claudia; Giulini, Maddalena; Gjelsten, Børge Kile; Gkaitatzis, Stamatios; Gkialas, Ioannis; Gkougkousis, Evangelos Leonidas; Gladilin, Leonid; Glasman, Claudia; Glatzer, Julian; Glaysher, Paul; Glazov, Alexandre; Goblirsch-Kolb, Maximilian; Goddard, Jack Robert; Godlewski, Jan; Goldfarb, Steven; Golling, Tobias; Golubkov, Dmitry; Gomes, Agostinho; Gonçalo, Ricardo; Goncalves Pinto Firmino Da Costa, Joao; Gonella, Laura; González de la Hoz, Santiago; Gonzalez Parra, Garoe; Gonzalez-Sevilla, Sergio; Goossens, Luc; Gorbounov, Petr Andreevich; Gordon, Howard; Gorelov, Igor; Gorini, Benedetto; Gorini, Edoardo; Gorišek, Andrej; Gornicki, Edward; Goshaw, Alfred; Gössling, Claus; Gostkin, Mikhail Ivanovitch; Goujdami, Driss; Goussiou, Anna; Govender, Nicolin; Grabas, Herve Marie Xavier; Graber, Lars; Grabowska-Bold, Iwona; Grafström, Per; Grahn, Karl-Johan; Gramling, Johanna; Gramstad, Eirik; Grancagnolo, Sergio; Grassi, Valerio; Gratchev, Vadim; Gray, Heather; Graziani, Enrico; Greenwood, Zeno Dixon; Gregersen, Kristian; Gregor, Ingrid-Maria; Grenier, Philippe; Griffiths, Justin; Grillo, Alexander; Grimm, Kathryn; Grinstein, Sebastian; Gris, Philippe Luc Yves; Grivaz, Jean-Francois; Grohs, Johannes Philipp; Grohsjean, Alexander; Gross, Eilam; Grosse-Knetter, Joern; Grossi, Giulio Cornelio; Grout, Zara Jane; Guan, Liang; Guenther, Jaroslav; Guescini, Francesco; Guest, Daniel; Gueta, Orel; Guido, Elisa; Guillemin, Thibault; Guindon, Stefan; Gul, Umar; Gumpert, Christian; Guo, Jun; Gupta, Shaun; Gutierrez, Phillip; Gutierrez Ortiz, Nicolas Gilberto; Gutschow, Christian; Guyot, Claude; Gwenlan, Claire; Gwilliam, Carl; Haas, Andy; Haber, Carl; Hadavand, Haleh Khani; Haddad, Nacim; Haefner, Petra; Hageböck, Stephan; Hajduk, Zbigniew; Hakobyan, Hrachya; Haleem, Mahsana; Haley, Joseph; Hall, David; Halladjian, Garabed; Hallewell, Gregory David; Hamacher, Klaus; Hamal, Petr; Hamano, Kenji; Hamer, Matthias; Hamilton, Andrew; Hamilton, Samuel; Hamity, Guillermo Nicolas; Hamnett, Phillip George; Han, Liang; Hanagaki, Kazunori; Hanawa, Keita; Hance, Michael; Hanke, Paul; Hanna, Remie; Hansen, Jørgen Beck; Hansen, Jorn Dines; Hansen, Maike Christina; Hansen, Peter Henrik; Hara, Kazuhiko; Hard, Andrew; Harenberg, Torsten; Hariri, Faten; Harkusha, Siarhei; Harrington, Robert; Harrison, Paul Fraser; Hartjes, Fred; Hasegawa, Makoto; Hasegawa, Satoshi; Hasegawa, Yoji; Hasib, A; Hassani, Samira; Haug, Sigve; Hauser, Reiner; Hauswald, Lorenz; Havranek, Miroslav; Hawkes, Christopher; Hawkings, Richard John; Hawkins, Anthony David; Hayashi, Takayasu; Hayden, Daniel; Hays, Chris; Hays, Jonathan Michael; Hayward, Helen; Haywood, Stephen; Head, Simon; Heck, Tobias; Hedberg, Vincent; Heelan, Louise; Heim, Sarah; Heim, Timon; Heinemann, Beate; Heinrich, Lukas; Hejbal, Jiri; Helary, Louis; Hellman, Sten; Hellmich, Dennis; Helsens, Clement; Henderson, James; Henderson, Robert; Heng, Yang; Hengler, Christopher; Henrichs, Anna; Henriques Correia, Ana Maria; Henrot-Versille, Sophie; Herbert, Geoffrey Henry; Hernández Jiménez, Yesenia; Herrberg-Schubert, Ruth; Herten, Gregor; Hertenberger, Ralf; Hervas, Luis; Hesketh, Gavin Grant; Hessey, Nigel; Hetherly, Jeffrey Wayne; Hickling, Robert; Higón-Rodriguez, Emilio; Hill, Ewan; Hill, John; Hiller, Karl Heinz; Hillier, Stephen; Hinchliffe, Ian; Hines, Elizabeth; Hinman, Rachel Reisner; Hirose, Minoru; Hirschbuehl, Dominic; Hobbs, John; Hod, Noam; Hodgkinson, Mark; Hodgson, Paul; Hoecker, Andreas; Hoeferkamp, Martin; Hoenig, Friedrich; Hohlfeld, Marc; Hohn, David; Holmes, Tova Ray; Homann, Michael; Hong, Tae Min; Hooft van Huysduynen, Loek; Hopkins, Walter; Horii, Yasuyuki; Horton, Arthur James; Hostachy, Jean-Yves; Hou, Suen; Hoummada, Abdeslam; Howard, Jacob; Howarth, James; Hrabovsky, Miroslav; Hristova, Ivana; Hrivnac, Julius; Hryn'ova, Tetiana; Hrynevich, Aliaksei; Hsu, Catherine; Hsu, Pai-hsien Jennifer; Hsu, Shih-Chieh; Hu, Diedi; Hu, Qipeng; Hu, Xueye; Huang, Yanping; Hubacek, Zdenek; Hubaut, Fabrice; Huegging, Fabian; Huffman, Todd Brian; Hughes, Emlyn; Hughes, Gareth; Huhtinen, Mika; Hülsing, Tobias Alexander; Huseynov, Nazim; Huston, Joey; Huth, John; Iacobucci, Giuseppe; Iakovidis, Georgios; Ibragimov, Iskander; Iconomidou-Fayard, Lydia; Ideal, Emma; Idrissi, Zineb; Iengo, Paolo; Igonkina, Olga; Iizawa, Tomoya; Ikegami, Yoichi; Ikematsu, Katsumasa; Ikeno, Masahiro; Ilchenko, Iurii; Iliadis, Dimitrios; Ilic, Nikolina; Inamaru, Yuki; Ince, Tayfun; Ioannou, Pavlos; Iodice, Mauro; Iordanidou, Kalliopi; Ippolito, Valerio; Irles Quiles, Adrian; Isaksson, Charlie; Ishino, Masaya; Ishitsuka, Masaki; Ishmukhametov, Renat; Issever, Cigdem; Istin, Serhat; Iturbe Ponce, Julia Mariana; Iuppa, Roberto; Ivarsson, Jenny; Iwanski, Wieslaw; Iwasaki, Hiroyuki; Izen, Joseph; Izzo, Vincenzo; Jabbar, Samina; Jackson, Brett; Jackson, Matthew; Jackson, Paul; Jaekel, Martin; Jain, Vivek; Jakobs, Karl; Jakobsen, Sune; Jakoubek, Tomas; Jakubek, Jan; Jamin, David Olivier; Jana, Dilip; Jansen, Eric; Jansky, Roland; Janssen, Jens; Janus, Michel; Jarlskog, Göran; Javadov, Namig; Javůrek, Tomáš; Jeanty, Laura; Jejelava, Juansher; Jeng, Geng-yuan; Jennens, David; Jenni, Peter; Jentzsch, Jennifer; Jeske, Carl; Jézéquel, Stéphane; Ji, Haoshuang; Jia, Jiangyong; Jiang, Yi; Jiggins, Stephen; Jimenez Pena, Javier; Jin, Shan; Jinaru, Adam; Jinnouchi, Osamu; Joergensen, Morten Dam; Johansson, Per; Johns, Kenneth; Jon-And, Kerstin; Jones, Graham; Jones, Roger; Jones, Tim; Jongmanns, Jan; Jorge, Pedro; Joshi, Kiran Daniel; Jovicevic, Jelena; Ju, Xiangyang; Jung, Christian; Jussel, Patrick; Juste Rozas, Aurelio; Kaci, Mohammed; Kaczmarska, Anna; Kado, Marumi; Kagan, Harris; Kagan, Michael; Kahn, Sebastien Jonathan; Kajomovitz, Enrique; Kalderon, Charles William; Kama, Sami; Kamenshchikov, Andrey; Kanaya, Naoko; Kaneda, Michiru; Kaneti, Steven; Kantserov, Vadim; Kanzaki, Junichi; Kaplan, Benjamin; Kapliy, Anton; Kar, Deepak; Karakostas, Konstantinos; Karamaoun, Andrew; Karastathis, Nikolaos; Kareem, Mohammad Jawad; Karnevskiy, Mikhail; Karpov, Sergey; Karpova, Zoya; Karthik, Krishnaiyengar; Kartvelishvili, Vakhtang; Karyukhin, Andrey; Kashif, Lashkar; Kass, Richard; Kastanas, Alex; Kataoka, Yousuke; Katre, Akshay; Katzy, Judith; Kawagoe, Kiyotomo; Kawamoto, Tatsuo; Kawamura, Gen; Kazama, Shingo; Kazanin, Vassili; Kazarinov, Makhail; Keeler, Richard; Kehoe, Robert; Keller, John; Kempster, Jacob Julian; Keoshkerian, Houry; Kepka, Oldrich; Kerševan, Borut Paul; Kersten, Susanne; Keyes, Robert; Khalil-zada, Farkhad; Khandanyan, Hovhannes; Khanov, Alexander; Kharlamov, Alexey; Khoo, Teng Jian; Khovanskiy, Valery; Khramov, Evgeniy; Khubua, Jemal; Kim, Hee Yeun; Kim, Hyeon Jin; Kim, Shinhong; Kim, Young-Kee; Kimura, Naoki; Kind, Oliver Maria; King, Barry; King, Matthew; King, Robert Steven Beaufoy; King, Samuel Burton; Kirk, Julie; Kiryunin, Andrey; Kishimoto, Tomoe; Kisielewska, Danuta; Kiss, Florian; Kiuchi, Kenji; Kivernyk, Oleh; Kladiva, Eduard; Klein, Matthew Henry; Klein, Max; Klein, Uta; Kleinknecht, Konrad; Klimek, Pawel; Klimentov, Alexei; Klingenberg, Reiner; Klinger, Joel Alexander; Klioutchnikova, Tatiana; Kluge, Eike-Erik; Kluit, Peter; Kluth, Stefan; Kneringer, Emmerich; Knoops, Edith; Knue, Andrea; Kobayashi, Aine; Kobayashi, Dai; Kobayashi, Tomio; Kobel, Michael; Kocian, Martin; Kodys, Peter; Koffas, Thomas; Koffeman, Els; Kogan, Lucy Anne; Kohlmann, Simon; Kohout, Zdenek; Kohriki, Takashi; Koi, Tatsumi; Kolanoski, Hermann; Koletsou, Iro; Komar, Aston; Komori, Yuto; Kondo, Takahiko; Kondrashova, Nataliia; Köneke, Karsten; König, Adriaan; König, Sebastian; Kono, Takanori; Konoplich, Rostislav; Konstantinidis, Nikolaos; Kopeliansky, Revital; Koperny, Stefan; Köpke, Lutz; Kopp, Anna Katharina; Korcyl, Krzysztof; Kordas, Kostantinos; Korn, Andreas; Korol, Aleksandr; Korolkov, Ilya; Korolkova, Elena; Kortner, Oliver; Kortner, Sandra; Kosek, Tomas; Kostyukhin, Vadim; Kotov, Vladislav; Kotwal, Ashutosh; Kourkoumeli-Charalampidi, Athina; Kourkoumelis, Christine; Kouskoura, Vasiliki; Koutsman, Alex; Kowalewski, Robert Victor; Kowalski, Tadeusz; Kozanecki, Witold; Kozhin, Anatoly; Kramarenko, Viktor; Kramberger, Gregor; Krasnopevtsev, Dimitriy; Krasny, Mieczyslaw Witold; Krasznahorkay, Attila; Kraus, Jana; Kravchenko, Anton; Kreiss, Sven; Kretz, Moritz; Kretzschmar, Jan; Kreutzfeldt, Kristof; Krieger, Peter; Krizka, Karol; Kroeninger, Kevin; Kroha, Hubert; Kroll, Joe; Kroseberg, Juergen; Krstic, Jelena; Kruchonak, Uladzimir; Krüger, Hans; Krumnack, Nils; Krumshteyn, Zinovii; Kruse, Amanda; Kruse, Mark; Kruskal, Michael; Kubota, Takashi; Kucuk, Hilal; Kuday, Sinan; Kuehn, Susanne; Kugel, Andreas; Kuger, Fabian; Kuhl, Andrew; Kuhl, Thorsten; Kukhtin, Victor; Kulchitsky, Yuri; Kuleshov, Sergey; Kuna, Marine; Kunigo, Takuto; Kupco, Alexander; Kurashige, Hisaya; Kurochkin, Yurii; Kurumida, Rie; Kus, Vlastimil; Kuwertz, Emma Sian; Kuze, Masahiro; Kvita, Jiri; Kwan, Tony; Kyriazopoulos, Dimitrios; La Rosa, Alessandro; La Rosa Navarro, Jose Luis; La Rotonda, Laura; Lacasta, Carlos; Lacava, Francesco; Lacey, James; Lacker, Heiko; Lacour, Didier; Lacuesta, Vicente Ramón; Ladygin, Evgueni; Lafaye, Remi; Laforge, Bertrand; Lagouri, Theodota; Lai, Stanley; Lambourne, Luke; Lammers, Sabine; Lampen, Caleb; Lampl, Walter; Lançon, Eric; Landgraf, Ulrich; Landon, Murrough; Lang, Valerie Susanne; Lange, J örn Christian; Lankford, Andrew; Lanni, Francesco; Lantzsch, Kerstin; Laplace, Sandrine; Lapoire, Cecile; Laporte, Jean-Francois; Lari, Tommaso; Lasagni Manghi, Federico; Lassnig, Mario; Laurelli, Paolo; Lavrijsen, Wim; Law, Alexander; Laycock, Paul; Le Dortz, Olivier; Le Guirriec, Emmanuel; Le Menedeu, Eve; LeBlanc, Matthew Edgar; LeCompte, Thomas; Ledroit-Guillon, Fabienne Agnes Marie; Lee, Claire Alexandra; Lee, Shih-Chang; Lee, Lawrence; Lefebvre, Guillaume; Lefebvre, Michel; Legger, Federica; Leggett, Charles; Lehan, Allan; Lehmann Miotto, Giovanna; Lei, Xiaowen; Leight, William Axel; Leisos, Antonios; Leister, Andrew Gerard; Leite, Marco Aurelio Lisboa; Leitner, Rupert; Lellouch, Daniel; Lemmer, Boris; Leney, Katharine; Lenz, Tatjana; Lenzi, Bruno; Leone, Robert; Leone, Sandra; Leonidopoulos, Christos; Leontsinis, Stefanos; Leroy, Claude; Lester, Christopher; Levchenko, Mikhail; Levêque, Jessica; Levin, Daniel; Levinson, Lorne; Levy, Mark; Lewis, Adrian; Leyko, Agnieszka; Leyton, Michael; Li, Bing; Li, Haifeng; Li, Ho Ling; Li, Lei; Li, Liang; Li, Shu; Li, Yichen; Liang, Zhijun; Liao, Hongbo; Liberti, Barbara; Liblong, Aaron; Lichard, Peter; Lie, Ki; Liebal, Jessica; Liebig, Wolfgang; Limbach, Christian; Limosani, Antonio; Lin, Simon; Lin, Tai-Hua; Linde, Frank; Lindquist, Brian Edward; Linnemann, James; Lipeles, Elliot; Lipniacka, Anna; Lisovyi, Mykhailo; Liss, Tony; Lissauer, David; Lister, Alison; Litke, Alan; Liu, Bo; Liu, Dong; Liu, Jian; Liu, Jianbei; Liu, Kun; Liu, Lulu; Liu, Miaoyuan; Liu, Minghui; Liu, Yanwen; Livan, Michele; Lleres, Annick; Llorente Merino, Javier; Lloyd, Stephen; Lo Sterzo, Francesco; Lobodzinska, Ewelina; Loch, Peter; Lockman, William; Loebinger, Fred; Loevschall-Jensen, Ask Emil; Loginov, Andrey; Lohse, Thomas; Lohwasser, Kristin; Lokajicek, Milos; Long, Brian Alexander; Long, Jonathan; Long, Robin Eamonn; Looper, Kristina Anne; Lopes, Lourenco; Lopez Mateos, David; Lopez Paredes, Brais; Lopez Paz, Ivan; Lorenz, Jeanette; Lorenzo Martinez, Narei; Losada, Marta; Loscutoff, Peter; Lösel, Philipp Jonathan; Lou, XinChou; Lounis, Abdenour; Love, Jeremy; Love, Peter; Lu, Nan; Lubatti, Henry; Luci, Claudio; Lucotte, Arnaud; Luehring, Frederick; Lukas, Wolfgang; Luminari, Lamberto; Lundberg, Olof; Lund-Jensen, Bengt; Lynn, David; Lysak, Roman; Lytken, Else; Ma, Hong; Ma, Lian Liang; Maccarrone, Giovanni; Macchiolo, Anna; Macdonald, Calum Michael; Machado Miguens, Joana; Macina, Daniela; Madaffari, Daniele; Madar, Romain; Maddocks, Harvey Jonathan; Mader, Wolfgang; Madsen, Alexander; Maeland, Steffen; Maeno, Tadashi; Maevskiy, Artem; Magradze, Erekle; Mahboubi, Kambiz; Mahlstedt, Joern; Maiani, Camilla; Maidantchik, Carmen; Maier, Andreas Alexander; Maier, Thomas; Maio, Amélia; Majewski, Stephanie; Makida, Yasuhiro; Makovec, Nikola; Malaescu, Bogdan; Malecki, Pawel; Maleev, Victor; Malek, Fairouz; Mallik, Usha; Malon, David; Malone, Caitlin; Maltezos, Stavros; Malyshev, Vladimir; Malyukov, Sergei; Mamuzic, Judita; Mancini, Giada; Mandelli, Beatrice; Mandelli, Luciano; Mandić, Igor; Mandrysch, Rocco; Maneira, José; Manfredini, Alessandro; Manhaes de Andrade Filho, Luciano; Manjarres Ramos, Joany; Mann, Alexander; Manning, Peter; Manousakis-Katsikakis, Arkadios; Mansoulie, Bruno; Mantifel, Rodger; Mantoani, Matteo; Mapelli, Livio; March, Luis; Marchiori, Giovanni; Marcisovsky, Michal; Marino, Christopher; Marjanovic, Marija; Marroquim, Fernando; Marsden, Stephen Philip; Marshall, Zach; Marti, Lukas Fritz; Marti-Garcia, Salvador; Martin, Brian Thomas; Martin, Tim; Martin, Victoria Jane; Martin dit Latour, Bertrand; Martinez, Mario; Martin-Haugh, Stewart; Martoiu, Victor Sorin; Martyniuk, Alex; Marx, Marilyn; Marzano, Francesco; Marzin, Antoine; Masetti, Lucia; Mashimo, Tetsuro; Mashinistov, Ruslan; Masik, Jiri; Maslennikov, Alexey; Massa, Ignazio; Massa, Lorenzo; Massol, Nicolas; Mastrandrea, Paolo; Mastroberardino, Anna; Masubuchi, Tatsuya; Mättig, Peter; Mattmann, Johannes; Maurer, Julien; Maxfield, Stephen; Maximov, Dmitriy; Mazini, Rachid; Mazza, Simone Michele; Mazzaferro, Luca; Mc Goldrick, Garrin; Mc Kee, Shawn Patrick; McCarn, Allison; McCarthy, Robert; McCarthy, Tom; McCubbin, Norman; McFarlane, Kenneth; Mcfayden, Josh; Mchedlidze, Gvantsa; McMahon, Steve; McPherson, Robert; Medinnis, Michael; Meehan, Samuel; Mehlhase, Sascha; Mehta, Andrew; Meier, Karlheinz; Meineck, Christian; Meirose, Bernhard; Mellado Garcia, Bruce Rafael; Meloni, Federico; Mengarelli, Alberto; Menke, Sven; Meoni, Evelin; Mercurio, Kevin Michael; Mergelmeyer, Sebastian; Mermod, Philippe; Merola, Leonardo; Meroni, Chiara; Merritt, Frank; Messina, Andrea; Metcalfe, Jessica; Mete, Alaettin Serhan; Meyer, Carsten; Meyer, Christopher; Meyer, Jean-Pierre; Meyer, Jochen; Middleton, Robin; Miglioranzi, Silvia; Mijović, Liza; Mikenberg, Giora; Mikestikova, Marcela; Mikuž, Marko; Milesi, Marco; Milic, Adriana; Miller, David; Mills, Corrinne; Milov, Alexander; Milstead, David; Minaenko, Andrey; Minami, Yuto; Minashvili, Irakli; Mincer, Allen; Mindur, Bartosz; Mineev, Mikhail; Ming, Yao; Mir, Lluisa-Maria; Mitani, Takashi; Mitrevski, Jovan; Mitsou, Vasiliki A; Miucci, Antonio; Miyagawa, Paul; Mjörnmark, Jan-Ulf; Moa, Torbjoern; Mochizuki, Kazuya; Mohapatra, Soumya; Mohr, Wolfgang; Molander, Simon; Moles-Valls, Regina; Mönig, Klaus; Monini, Caterina; Monk, James; Monnier, Emmanuel; Montejo Berlingen, Javier; Monticelli, Fernando; Monzani, Simone; Moore, Roger; Morange, Nicolas; Moreno, Deywis; Moreno Llácer, María; Morettini, Paolo; Morgenstern, Marcus; Morii, Masahiro; Morinaga, Masahiro; Morisbak, Vanja; Moritz, Sebastian; Morley, Anthony Keith; Mornacchi, Giuseppe; Morris, John; Mortensen, Simon Stark; Morton, Alexander; Morvaj, Ljiljana; Mosidze, Maia; Moss, Josh; Motohashi, Kazuki; Mount, Richard; Mountricha, Eleni; Mouraviev, Sergei; Moyse, Edward; Muanza, Steve; Mudd, Richard; Mueller, Felix; Mueller, James; Mueller, Klemens; Mueller, Ralph Soeren Peter; Mueller, Thibaut; Muenstermann, Daniel; Mullen, Paul; Munwes, Yonathan; Murillo Quijada, Javier Alberto; Murray, Bill; Musheghyan, Haykuhi; Musto, Elisa; Myagkov, Alexey; Myska, Miroslav; Nackenhorst, Olaf; Nadal, Jordi; Nagai, Koichi; Nagai, Ryo; Nagai, Yoshikazu; Nagano, Kunihiro; Nagarkar, Advait; Nagasaka, Yasushi; Nagata, Kazuki; Nagel, Martin; Nagy, Elemer; Nairz, Armin Michael; Nakahama, Yu; Nakamura, Koji; Nakamura, Tomoaki; Nakano, Itsuo; Namasivayam, Harisankar; Naranjo Garcia, Roger Felipe; Narayan, Rohin; Naumann, Thomas; Navarro, Gabriela; Nayyar, Ruchika; Neal, Homer; Nechaeva, Polina; Neep, Thomas James; Nef, Pascal Daniel; Negri, Andrea; Negrini, Matteo; Nektarijevic, Snezana; Nellist, Clara; Nelson, Andrew; Nemecek, Stanislav; Nemethy, Peter; Nepomuceno, Andre Asevedo; Nessi, Marzio; Neubauer, Mark; Neumann, Manuel; Neves, Ricardo; Nevski, Pavel; Newman, Paul; Nguyen, Duong Hai; Nickerson, Richard; Nicolaidou, Rosy; Nicquevert, Bertrand; Nielsen, Jason; Nikiforou, Nikiforos; Nikiforov, Andriy; Nikolaenko, Vladimir; Nikolic-Audit, Irena; Nikolopoulos, Konstantinos; Nilsen, Jon Kerr; Nilsson, Paul; Ninomiya, Yoichi; Nisati, Aleandro; Nisius, Richard; Nobe, Takuya; Nomachi, Masaharu; Nomidis, Ioannis; Nooney, Tamsin; Norberg, Scarlet; Nordberg, Markus; Novgorodova, Olga; Nowak, Sebastian; Nozaki, Mitsuaki; Nozka, Libor; Ntekas, Konstantinos; Nunes Hanninger, Guilherme; Nunnemann, Thomas; Nurse, Emily; Nuti, Francesco; O'Brien, Brendan Joseph; O'grady, Fionnbarr; O'Neil, Dugan; O'Shea, Val; Oakham, Gerald; Oberlack, Horst; Obermann, Theresa; Ocariz, Jose; Ochi, Atsuhiko; Ochoa, Ines; Ochoa-Ricoux, Juan Pedro; Oda, Susumu; Odaka, Shigeru; Ogren, Harold; Oh, Alexander; Oh, Seog; Ohm, Christian; Ohman, Henrik; Oide, Hideyuki; Okamura, Wataru; Okawa, Hideki; Okumura, Yasuyuki; Okuyama, Toyonobu; Olariu, Albert; Olivares Pino, Sebastian Andres; Oliveira Damazio, Denis; Oliver Garcia, Elena; Olszewski, Andrzej; Olszowska, Jolanta; Onofre, António; Onyisi, Peter; Oram, Christopher; Oreglia, Mark; Oren, Yona; Orestano, Domizia; Orlando, Nicola; Oropeza Barrera, Cristina; Orr, Robert; Osculati, Bianca; Ospanov, Rustem; Otero y Garzon, Gustavo; Otono, Hidetoshi; Ouchrif, Mohamed; Ouellette, Eric; Ould-Saada, Farid; Ouraou, Ahmimed; Oussoren, Koen Pieter; Ouyang, Qun; Ovcharova, Ana; Owen, Mark; Owen, Rhys Edward; Ozcan, Veysi Erkcan; Ozturk, Nurcan; Pachal, Katherine; Pacheco Pages, Andres; Padilla Aranda, Cristobal; Pagáčová, Martina; Pagan Griso, Simone; Paganis, Efstathios; Pahl, Christoph; Paige, Frank; Pais, Preema; Pajchel, Katarina; Palacino, Gabriel; Palestini, Sandro; Palka, Marek; Pallin, Dominique; Palma, Alberto; Pan, Yibin; Panagiotopoulou, Evgenia; Pandini, Carlo Enrico; Panduro Vazquez, William; Pani, Priscilla; Panitkin, Sergey; Pantea, Dan; Paolozzi, Lorenzo; Papadopoulou, Theodora; Papageorgiou, Konstantinos; Paramonov, Alexander; Paredes Hernandez, Daniela; Parker, Michael Andrew; Parker, Kerry Ann; Parodi, Fabrizio; Parsons, John; Parzefall, Ulrich; Pasqualucci, Enrico; Passaggio, Stefano; Pastore, Fernanda; Pastore, Francesca; Pásztor, Gabriella; Pataraia, Sophio; Patel, Nikhul; Pater, Joleen; Pauly, Thilo; Pearce, James; Pearson, Benjamin; Pedersen, Lars Egholm; Pedersen, Maiken; Pedraza Lopez, Sebastian; Pedro, Rute; Peleganchuk, Sergey; Pelikan, Daniel; Peng, Haiping; Penning, Bjoern; Penwell, John; Perepelitsa, Dennis; Perez Codina, Estel; Pérez García-Estañ, María Teresa; Perini, Laura; Pernegger, Heinz; Perrella, Sabrina; Peschke, Richard; Peshekhonov, Vladimir; Peters, Krisztian; Peters, Yvonne; Petersen, Brian; Petersen, Troels; Petit, Elisabeth; Petridis, Andreas; Petridou, Chariclia; Petrolo, Emilio; Petrucci, Fabrizio; Pettersson, Nora Emilia; Pezoa, Raquel; Phillips, Peter William; Piacquadio, Giacinto; Pianori, Elisabetta; Picazio, Attilio; Piccaro, Elisa; Piccinini, Maurizio; Pickering, Mark Andrew; Piegaia, Ricardo; Pignotti, David; Pilcher, James; Pilkington, Andrew; Pina, João Antonio; Pinamonti, Michele; Pinfold, James; Pingel, Almut; Pinto, Belmiro; Pires, Sylvestre; Pitt, Michael; Pizio, Caterina; Plazak, Lukas; Pleier, Marc-Andre; Pleskot, Vojtech; Plotnikova, Elena; Plucinski, Pawel; Pluth, Daniel; Poettgen, Ruth; Poggioli, Luc; Pohl, David-leon; Polesello, Giacomo; Policicchio, Antonio; Polifka, Richard; Polini, Alessandro; Pollard, Christopher Samuel; Polychronakos, Venetios; Pommès, Kathy; Pontecorvo, Ludovico; Pope, Bernard; Popeneciu, Gabriel Alexandru; Popovic, Dragan; Poppleton, Alan; Pospisil, Stanislav; Potamianos, Karolos; Potrap, Igor; Potter, Christina; Potter, Christopher; Poulard, Gilbert; Poveda, Joaquin; Pozdnyakov, Valery; Pralavorio, Pascal; Pranko, Aliaksandr; Prasad, Srivas; Prell, Soeren; Price, Darren; Price, Lawrence; Primavera, Margherita; Prince, Sebastien; Proissl, Manuel; Prokofiev, Kirill; Prokoshin, Fedor; Protopapadaki, Eftychia-sofia; Protopopescu, Serban; Proudfoot, James; Przybycien, Mariusz; Ptacek, Elizabeth; Puddu, Daniele; Pueschel, Elisa; Puldon, David; Purohit, Milind; Puzo, Patrick; Qian, Jianming; Qin, Gang; Qin, Yang; Quadt, Arnulf; Quarrie, David; Quayle, William; Queitsch-Maitland, Michaela; Quilty, Donnchadha; Raddum, Silje; Radeka, Veljko; Radescu, Voica; Radhakrishnan, Sooraj Krishnan; Radloff, Peter; Rados, Pere; Ragusa, Francesco; Rahal, Ghita; Rajagopalan, Srinivasan; Rammensee, Michael; Rangel-Smith, Camila; Rauscher, Felix; Rave, Stefan; Ravenscroft, Thomas; Raymond, Michel; Read, Alexander Lincoln; Readioff, Nathan Peter; Rebuzzi, Daniela; Redelbach, Andreas; Redlinger, George; Reece, Ryan; Reeves, Kendall; Rehnisch, Laura; Reisin, Hernan; Relich, Matthew; Rembser, Christoph; Ren, Huan; Renaud, Adrien; Rescigno, Marco; Resconi, Silvia; Rezanova, Olga; Reznicek, Pavel; Rezvani, Reyhaneh; Richter, Robert; Richter, Stefan; Richter-Was, Elzbieta; Ricken, Oliver; Ridel, Melissa; Rieck, Patrick; Riegel, Christian Johann; Rieger, Julia; Rijssenbeek, Michael; Rimoldi, Adele; Rinaldi, Lorenzo; Ristić, Branislav; Ritsch, Elmar; Riu, Imma; Rizatdinova, Flera; Rizvi, Eram; Robertson, Steven; Robichaud-Veronneau, Andree; Robinson, Dave; Robinson, James; Robson, Aidan; Roda, Chiara; Roe, Shaun; Røhne, Ole; Rolli, Simona; Romaniouk, Anatoli; Romano, Marino; Romano Saez, Silvestre Marino; Romero Adam, Elena; Rompotis, Nikolaos; Ronzani, Manfredi; Roos, Lydia; Ros, Eduardo; Rosati, Stefano; Rosbach, Kilian; Rose, Peyton; Rosendahl, Peter Lundgaard; Rosenthal, Oliver; Rossetti, Valerio; Rossi, Elvira; Rossi, Leonardo Paolo; Rosten, Rachel; Rotaru, Marina; Roth, Itamar; Rothberg, Joseph; Rousseau, David; Royon, Christophe; Rozanov, Alexandre; Rozen, Yoram; Ruan, Xifeng; Rubbo, Francesco; Rubinskiy, Igor; Rud, Viacheslav; Rudolph, Christian; Rudolph, Matthew Scott; Rühr, Frederik; Ruiz-Martinez, Aranzazu; Rurikova, Zuzana; Rusakovich, Nikolai; Ruschke, Alexander; Russell, Heather; Rutherfoord, John; Ruthmann, Nils; Ryabov, Yury; Rybar, Martin; Rybkin, Grigori; Ryder, Nick; Saavedra, Aldo; Sabato, Gabriele; Sacerdoti, Sabrina; Saddique, Asif; Sadrozinski, Hartmut; Sadykov, Renat; Safai Tehrani, Francesco; Saimpert, Matthias; Sakamoto, Hiroshi; Sakurai, Yuki; Salamanna, Giuseppe; Salamon, Andrea; Saleem, Muhammad; Salek, David; Sales De Bruin, Pedro Henrique; Salihagic, Denis; Salnikov, Andrei; Salt, José; Salvatore, Daniela; Salvatore, Pasquale Fabrizio; Salvucci, Antonio; Salzburger, Andreas; Sampsonidis, Dimitrios; Sanchez, Arturo; Sánchez, Javier; Sanchez Martinez, Victoria; Sandaker, Heidi; Sandbach, Ruth Laura; Sander, Heinz Georg; Sanders, Michiel; Sandhoff, Marisa; Sandoval, Carlos; Sandstroem, Rikard; Sankey, Dave; Sannino, Mario; Sansoni, Andrea; Santoni, Claudio; Santonico, Rinaldo; Santos, Helena; Santoyo Castillo, Itzebelt; Sapp, Kevin; Sapronov, Andrey; Saraiva, João; Sarrazin, Bjorn; Sasaki, Osamu; Sasaki, Yuichi; Sato, Koji; Sauvage, Gilles; Sauvan, Emmanuel; Savage, Graham; Savard, Pierre; Sawyer, Craig; Sawyer, Lee; Saxon, James; Sbarra, Carla; Sbrizzi, Antonio; Scanlon, Tim; Scannicchio, Diana; Scarcella, Mark; Scarfone, Valerio; Schaarschmidt, Jana; Schacht, Peter; Schaefer, Douglas; Schaefer, Ralph; Schaeffer, Jan; Schaepe, Steffen; Schaetzel, Sebastian; Schäfer, Uli; Schaffer, Arthur; Schaile, Dorothee; Schamberger, R~Dean; Scharf, Veit; Schegelsky, Valery; Scheirich, Daniel; Schernau, Michael; Schiavi, Carlo; Schillo, Christian; Schioppa, Marco; Schlenker, Stefan; Schmidt, Evelyn; Schmieden, Kristof; Schmitt, Christian; Schmitt, Sebastian; Schmitt, Stefan; Schneider, Basil; Schnellbach, Yan Jie; Schnoor, Ulrike; Schoeffel, Laurent; Schoening, Andre; Schoenrock, Bradley Daniel; Schopf, Elisabeth; Schorlemmer, Andre Lukas; Schott, Matthias; Schouten, Doug; Schovancova, Jaroslava; Schramm, Steven; Schreyer, Manuel; Schroeder, Christian; Schuh, Natascha; Schultens, Martin Johannes; Schultz-Coulon, Hans-Christian; Schulz, Holger; Schumacher, Markus; Schumm, Bruce; Schune, Philippe; Schwanenberger, Christian; Schwartzman, Ariel; Schwarz, Thomas Andrew; Schwegler, Philipp; Schwemling, Philippe; Schwienhorst, Reinhard; Schwindling, Jerome; Schwindt, Thomas; Schwoerer, Maud; Sciacca, Gianfranco; Scifo, Estelle; Sciolla, Gabriella; Scuri, Fabrizio; Scutti, Federico; Searcy, Jacob; Sedov, George; Sedykh, Evgeny; Seema, Pienpen; Seidel, Sally; Seiden, Abraham; Seifert, Frank; Seixas, José; Sekhniaidze, Givi; Sekhon, Karishma; Sekula, Stephen; Selbach, Karoline Elfriede; Seliverstov, Dmitry; Semprini-Cesari, Nicola; Serfon, Cedric; Serin, Laurent; Serkin, Leonid; Serre, Thomas; Sessa, Marco; Seuster, Rolf; Severini, Horst; Sfiligoj, Tina; Sforza, Federico; Sfyrla, Anna; Shabalina, Elizaveta; Shamim, Mansoora; Shan, Lianyou; Shang, Ruo-yu; Shank, James; Shapiro, Marjorie; Shatalov, Pavel; Shaw, Kate; Shaw, Savanna Marie; Shcherbakova, Anna; Shehu, Ciwake Yusufu; Sherwood, Peter; Shi, Liaoshan; Shimizu, Shima; Shimmin, Chase Owen; Shimojima, Makoto; Shiyakova, Mariya; Shmeleva, Alevtina; Shoaleh Saadi, Diane; Shochet, Mel; Shojaii, Seyedruhollah; Shrestha, Suyog; Shulga, Evgeny; Shupe, Michael; Shushkevich, Stanislav; Sicho, Petr; Sidiropoulou, Ourania; Sidorov, Dmitri; Sidoti, Antonio; Siegert, Frank; Sijacki, Djordje; Silva, José; Silver, Yiftah; Silverstein, Samuel; Simak, Vladislav; Simard, Olivier; Simic, Ljiljana; Simion, Stefan; Simioni, Eduard; Simmons, Brinick; Simon, Dorian; Simoniello, Rosa; Sinervo, Pekka; Sinev, Nikolai; Siragusa, Giovanni; Sisakyan, Alexei; Sivoklokov, Serguei; Sjölin, Jörgen; Sjursen, Therese; Skinner, Malcolm Bruce; Skottowe, Hugh Philip; Skubic, Patrick; Slater, Mark; Slavicek, Tomas; Slawinska, Magdalena; Sliwa, Krzysztof; Smakhtin, Vladimir; Smart, Ben; Smestad, Lillian; Smirnov, Sergei; Smirnov, Yury; Smirnova, Lidia; Smirnova, Oxana; Smith, Matthew; Smizanska, Maria; Smolek, Karel; Snesarev, Andrei; Snidero, Giacomo; Snyder, Scott; Sobie, Randall; Socher, Felix; Soffer, Abner; Soh, Dart-yin; Solans, Carlos; Solar, Michael; Solc, Jaroslav; Soldatov, Evgeny; Soldevila, Urmila; Solodkov, Alexander; Soloshenko, Alexei; Solovyanov, Oleg; Solovyev, Victor; Sommer, Philip; Song, Hong Ye; Soni, Nitesh; Sood, Alexander; Sopczak, Andre; Sopko, Bruno; Sopko, Vit; Sorin, Veronica; Sosa, David; Sosebee, Mark; Sotiropoulou, Calliope Louisa; Soualah, Rachik; Soueid, Paul; Soukharev, Andrey; South, David; Spagnolo, Stefania; Spalla, Margherita; Spanò, Francesco; Spearman, William Robert; Spettel, Fabian; Spighi, Roberto; Spigo, Giancarlo; Spiller, Laurence Anthony; Spousta, Martin; Spreitzer, Teresa; St Denis, Richard Dante; Staerz, Steffen; Stahlman, Jonathan; Stamen, Rainer; Stamm, Soren; Stanecka, Ewa; Stanescu, Cristian; Stanescu-Bellu, Madalina; Stanitzki, Marcel Michael; Stapnes, Steinar; Starchenko, Evgeny; Stark, Jan; Staroba, Pavel; Starovoitov, Pavel; Staszewski, Rafal; Stavina, Pavel; Steinberg, Peter; Stelzer, Bernd; Stelzer, Harald Joerg; Stelzer-Chilton, Oliver; Stenzel, Hasko; Stern, Sebastian; Stewart, Graeme; Stillings, Jan Andre; Stockton, Mark; Stoebe, Michael; Stoicea, Gabriel; Stolte, Philipp; Stonjek, Stefan; Stradling, Alden; Straessner, Arno; Stramaglia, Maria Elena; Strandberg, Jonas; Strandberg, Sara; Strandlie, Are; Strauss, Emanuel; Strauss, Michael; Strizenec, Pavol; Ströhmer, Raimund; Strom, David; Stroynowski, Ryszard; Strubig, Antonia; Stucci, Stefania Antonia; Stugu, Bjarne; Styles, Nicholas Adam; Su, Dong; Su, Jun; Subramaniam, Rajivalochan; Succurro, Antonella; Sugaya, Yorihito; Suhr, Chad; Suk, Michal; Sulin, Vladimir; Sultansoy, Saleh; Sumida, Toshi; Sun, Siyuan; Sun, Xiaohu; Sundermann, Jan Erik; Suruliz, Kerim; Susinno, Giancarlo; Sutton, Mark; Suzuki, Shota; Suzuki, Yu; Svatos, Michal; Swedish, Stephen; Swiatlowski, Maximilian; Sykora, Ivan; Sykora, Tomas; Ta, Duc; Taccini, Cecilia; Tackmann, Kerstin; Taenzer, Joe; Taffard, Anyes; Tafirout, Reda; Taiblum, Nimrod; Takai, Helio; Takashima, Ryuichi; Takeda, Hiroshi; Takeshita, Tohru; Takubo, Yosuke; Talby, Mossadek; Talyshev, Alexey; Tam, Jason; Tan, Kong Guan; Tanaka, Junichi; Tanaka, Reisaburo; Tanaka, Shuji; Tannenwald, Benjamin Bordy; Tannoury, Nancy; Tapprogge, Stefan; Tarem, Shlomit; Tarrade, Fabien; Tartarelli, Giuseppe Francesco; Tas, Petr; Tasevsky, Marek; Tashiro, Takuya; Tassi, Enrico; Tavares Delgado, Ademar; Tayalati, Yahya; Taylor, Frank; Taylor, Geoffrey; Taylor, Wendy; Teischinger, Florian Alfred; Teixeira Dias Castanheira, Matilde; Teixeira-Dias, Pedro; Temming, Kim Katrin; Ten Kate, Herman; Teng, Ping-Kun; Teoh, Jia Jian; Tepel, Fabian-Phillipp; Terada, Susumu; Terashi, Koji; Terron, Juan; Terzo, Stefano; Testa, Marianna; Teuscher, Richard; Therhaag, Jan; Theveneaux-Pelzer, Timothée; Thomas, Juergen; Thomas-Wilsker, Joshuha; Thompson, Emily; Thompson, Paul; Thompson, Ray; Thompson, Stan; Thomsen, Lotte Ansgaard; Thomson, Evelyn; Thomson, Mark; Thun, Rudolf; Tibbetts, Mark James; Ticse Torres, Royer Edson; Tikhomirov, Vladimir; Tikhonov, Yury; Timoshenko, Sergey; Tiouchichine, Elodie; Tipton, Paul; Tisserant, Sylvain; Todorov, Theodore; Todorova-Nova, Sharka; Tojo, Junji; Tokár, Stanislav; Tokushuku, Katsuo; Tollefson, Kirsten; Tolley, Emma; Tomlinson, Lee; Tomoto, Makoto; Tompkins, Lauren; Toms, Konstantin; Torrence, Eric; Torres, Heberth; Torró Pastor, Emma; Toth, Jozsef; Touchard, Francois; Tovey, Daniel; Trefzger, Thomas; Tremblet, Louis; Tricoli, Alessandro; Trigger, Isabel Marian; Trincaz-Duvoid, Sophie; Tripiana, Martin; Trischuk, William; Trocmé, Benjamin; Troncon, Clara; Trottier-McDonald, Michel; Trovatelli, Monica; True, Patrick; Truong, Loan; Trzebinski, Maciej; Trzupek, Adam; Tsarouchas, Charilaos; Tseng, Jeffrey; Tsiareshka, Pavel; Tsionou, Dimitra; Tsipolitis, Georgios; Tsirintanis, Nikolaos; Tsiskaridze, Shota; Tsiskaridze, Vakhtang; Tskhadadze, Edisher; Tsukerman, Ilya; Tsulaia, Vakhtang; Tsuno, Soshi; Tsybychev, Dmitri; Tudorache, Alexandra; Tudorache, Valentina; Tuna, Alexander Naip; Tupputi, Salvatore; Turchikhin, Semen; Turecek, Daniel; Turra, Ruggero; Turvey, Andrew John; Tuts, Michael; Tykhonov, Andrii; Tylmad, Maja; Tyndel, Mike; Ueda, Ikuo; Ueno, Ryuichi; Ughetto, Michael; Ugland, Maren; Uhlenbrock, Mathias; Ukegawa, Fumihiko; Unal, Guillaume; Undrus, Alexander; Unel, Gokhan; Ungaro, Francesca; Unno, Yoshinobu; Unverdorben, Christopher; Urban, Jozef; Urquijo, Phillip; Urrejola, Pedro; Usai, Giulio; Usanova, Anna; Vacavant, Laurent; Vacek, Vaclav; Vachon, Brigitte; Valderanis, Chrysostomos; Valencic, Nika; Valentinetti, Sara; Valero, Alberto; Valery, Loic; Valkar, Stefan; Valladolid Gallego, Eva; Vallecorsa, Sofia; Valls Ferrer, Juan Antonio; Van Den Wollenberg, Wouter; Van Der Deijl, Pieter; van der Geer, Rogier; van der Graaf, Harry; Van Der Leeuw, Robin; van Eldik, Niels; van Gemmeren, Peter; Van Nieuwkoop, Jacobus; van Vulpen, Ivo; van Woerden, Marius Cornelis; Vanadia, Marco; Vandelli, Wainer; Vanguri, Rami; Vaniachine, Alexandre; Vannucci, Francois; Vardanyan, Gagik; Vari, Riccardo; Varnes, Erich; Varol, Tulin; Varouchas, Dimitris; Vartapetian, Armen; Varvell, Kevin; Vazeille, Francois; Vazquez Schroeder, Tamara; Veatch, Jason; Veloso, Filipe; Velz, Thomas; Veneziano, Stefano; Ventura, Andrea; Ventura, Daniel; Venturi, Manuela; Venturi, Nicola; Venturini, Alessio; Vercesi, Valerio; Verducci, Monica; Verkerke, Wouter; Vermeulen, Jos; Vest, Anja; Vetterli, Michel; Viazlo, Oleksandr; Vichou, Irene; Vickey, Trevor; Vickey Boeriu, Oana Elena; Viehhauser, Georg; Viel, Simon; Vigne, Ralph; Villa, Mauro; Villaplana Perez, Miguel; Vilucchi, Elisabetta; Vincter, Manuella; Vinogradov, Vladimir; Vivarelli, Iacopo; Vives Vaque, Francesc; Vlachos, Sotirios; Vladoiu, Dan; Vlasak, Michal; Vogel, Marcelo; Vokac, Petr; Volpi, Guido; Volpi, Matteo; von der Schmitt, Hans; von Radziewski, Holger; von Toerne, Eckhard; Vorobel, Vit; Vorobev, Konstantin; Vos, Marcel; Voss, Rudiger; Vossebeld, Joost; Vranjes, Nenad; Vranjes Milosavljevic, Marija; Vrba, Vaclav; Vreeswijk, Marcel; Vuillermet, Raphael; Vukotic, Ilija; Vykydal, Zdenek; Wagner, Peter; Wagner, Wolfgang; Wahlberg, Hernan; Wahrmund, Sebastian; Wakabayashi, Jun; Walder, James; Walker, Rodney; Walkowiak, Wolfgang; Wang, Chao; Wang, Fuquan; Wang, Haichen; Wang, Hulin; Wang, Jike; Wang, Jin; Wang, Kuhan; Wang, Rui; Wang, Song-Ming; Wang, Tan; Wang, Xiaoxiao; Wanotayaroj, Chaowaroj; Warburton, Andreas; Ward, Patricia; Wardrope, David Robert; Warsinsky, Markus; Washbrook, Andrew; Wasicki, Christoph; Watkins, Peter; Watson, Alan; Watson, Ian; Watson, Miriam; Watts, Gordon; Watts, Stephen; Waugh, Ben; Webb, Samuel; Weber, Michele; Weber, Stefan Wolf; Webster, Jordan S; Weidberg, Anthony; Weinert, Benjamin; Weingarten, Jens; Weiser, Christian; Weits, Hartger; Wells, Phillippa; Wenaus, Torre; Wengler, Thorsten; Wenig, Siegfried; Wermes, Norbert; Werner, Matthias; Werner, Per; Wessels, Martin; Wetter, Jeffrey; Whalen, Kathleen; Wharton, Andrew Mark; White, Andrew; White, Martin; White, Ryan; White, Sebastian; Whiteson, Daniel; Wickens, Fred; Wiedenmann, Werner; Wielers, Monika; Wienemann, Peter; Wiglesworth, Craig; Wiik-Fuchs, Liv Antje Mari; Wildauer, Andreas; Wilkens, Henric George; Williams, Hugh; Williams, Sarah; Willis, Christopher; Willocq, Stephane; Wilson, Alan; Wilson, John; Wingerter-Seez, Isabelle; Winklmeier, Frank; Winter, Benedict Tobias; Wittgen, Matthias; Wittkowski, Josephine; Wollstadt, Simon Jakob; Wolter, Marcin Wladyslaw; Wolters, Helmut; Wosiek, Barbara; Wotschack, Jorg; Woudstra, Martin; Wozniak, Krzysztof; Wu, Mengqing; Wu, Miles; Wu, Sau Lan; Wu, Xin; Wu, Yusheng; Wyatt, Terry Richard; Wynne, Benjamin; Xella, Stefania; Xu, Da; Xu, Lailin; Yabsley, Bruce; Yacoob, Sahal; Yakabe, Ryota; Yamada, Miho; Yamaguchi, Yohei; Yamamoto, Akira; Yamamoto, Shimpei; Yamanaka, Takashi; Yamauchi, Katsuya; Yamazaki, Yuji; Yan, Zhen; Yang, Haijun; Yang, Hongtao; Yang, Yi; Yao, Liwen; Yao, Weiming; Yasu, Yoshiji; Yatsenko, Elena; Yau Wong, Kaven Henry; Ye, Jingbo; Ye, Shuwei; Yeletskikh, Ivan; Yen, Andy L; Yildirim, Eda; Yorita, Kohei; Yoshida, Rikutaro; Yoshihara, Keisuke; Young, Charles; Young, Christopher John; Youssef, Saul; Yu, David Ren-Hwa; Yu, Jaehoon; Yu, Jiaming; Yu, Jie; Yuan, Li; Yurkewicz, Adam; Yusuff, Imran; Zabinski, Bartlomiej; Zaidan, Remi; Zaitsev, Alexander; Zalieckas, Justas; Zaman, Aungshuman; Zambito, Stefano; Zanello, Lucia; Zanzi, Daniele; Zeitnitz, Christian; Zeman, Martin; Zemla, Andrzej; Zengel, Keith; Zenin, Oleg; Ženiš, Tibor; Zerwas, Dirk; Zhang, Dongliang; Zhang, Fangzhou; Zhang, Jinlong; Zhang, Lei; Zhang, Ruiqi; Zhang, Xueyao; Zhang, Zhiqing; Zhao, Xiandong; Zhao, Yongke; Zhao, Zhengguo; Zhemchugov, Alexey; Zhong, Jiahang; Zhou, Bing; Zhou, Chen; Zhou, Lei; Zhou, Li; Zhou, Ning; Zhu, Cheng Guang; Zhu, Hongbo; Zhu, Junjie; Zhu, Yingchun; Zhuang, Xuai; Zhukov, Konstantin; Zibell, Andre; Zieminska, Daria; Zimine, Nikolai; Zimmermann, Christoph; Zimmermann, Stephanie; Zinonos, Zinonas; Zinser, Markus; Ziolkowski, Michael; Živković, Lidija; Zobernig, Georg; Zoccoli, Antonio; zur Nedden, Martin; Zurzolo, Giovanni; Zwalinski, Lukasz

    2015-08-27

    A search for Higgs boson production in association with a $W$ or $Z$ boson, in the $H \\rightarrow WW^{\\ast}$ decay channel, is performed with a data sample collected with the ATLAS detector at the LHC in proton-proton collisions at centre-of-mass energies $\\sqrt{s}=7$ TeV and 8 TeV, corresponding to integrated luminosities of 4.5 ${\\rm fb}^{-1}$ and 20.3 ${\\rm fb}^{-1}$, respectively. The $WH$ production mode is studied in two-lepton and three-lepton final states, while two-lepton and four-lepton final states are used to search for the $ZH$ production mode. The observed significance, for the combined $WH$ and $ZH$ production, is 2.5 standard deviations while a significance of 0.9 standard deviations is expected in the Standard Model Higgs boson hypothesis. The ratio of the combined $WH$ and $ZH$ signal yield to the Standard Model expectation, $\\mu_{VH}$, is found to be $\\mu_{VH} =3.0^{+1.3}_{-1.1}{\\, {(\\rm stat.)}}^{+1.0}_{-0.7}{\\,{(\\rm sys.)}}$ for the Higgs boson mass of 125.36 GeV. The $WH$ and $ZH$ produc...

  8. An economic analysis of stretch-out for Angra-1 reactor

    International Nuclear Information System (INIS)

    Sakai, M.

    1989-01-01

    An application of NUCOST code for calculating nuclear energy cost is presented. Ann optimization of stretch-out for Angra-1 reactor based on international costs of nuclear fuel, operation and maintenance is done. (M.C.K.)

  9. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  10. Welding electrode for peripheral welds of A-1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Lakatos, L.

    1975-01-01

    The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)

  11. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  12. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-01-01

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  13. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2003-07-01

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

  14. . Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

    International Nuclear Information System (INIS)

    Yusuf, I; Mati, A. A.

    2010-01-01

    The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

  15. Treatment of arsenic contaminated water in a batch reactor by using Ralstonia eutropha MTCC 2487 and granular activated carbon

    International Nuclear Information System (INIS)

    Mondal, P.; Majumder, C.B.; Mohanty, B.

    2008-01-01

    This paper presents the observations on the bio-removal of arsenic from contaminated water by using Ralstonia eutropha MTCC 2487 and activated carbon in a batch reactor. The effects of agitation time, pH, type of granular activated carbon (GAC) and initial arsenic concentration (As o ) on the % removal of arsenic have been discussed. Under the experimental conditions, optimum removal was obtained at the pH of 6-7 with agitation time of 100 h. The % removal of As(T) increased initially with the increase in As o and after attaining the maximum removal (∼86%) at the As o value of around 15 ppm, it started to decrease. Simultaneous adsorption bioaccumulation (SABA) was observed, when fresh GAC was used as supporting media for bacterial immobilization. In case of SABA, the % removal of As(III) was almost similar (only ∼1% more) to the additive values of individual removal of As(III) obtained by only adsorption and only bio-adsorption. However, for As(V) the % removal was less (∼8%) than the additive value of the individual % removals obtained by only adsorption and bio-adsorption. Percentage removal of Fe, Mn, Cu and Zn were 65.17%, 72.76%, 98.6% and 99.31%, respectively. Maximum regeneration (∼99.4%) of the used bio-adsorbent was achieved by the treatment with 5NH 2 SO 4 followed by 1N NaOH and 30% H 2 O 2 in HNO 3 . The fitness of the isotherms to predict the specific uptake for bio-adsorption/accumulation process has been found to decrease in the following order: Temkin isotherm > Langmuir isotherm > Freundlich isotherm. For the adsorption process with fresh GAC the corresponding order is Freundlich isotherm > Langmuir isotherm > Temkin isotherm for As(V) and As(T). However, for As(III) it was Langmuir > Temkin > Freundlich

  16. VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

    International Nuclear Information System (INIS)

    Matejka, K.; Sklenka, L.

    2003-01-01

    The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)

  17. FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

    International Nuclear Information System (INIS)

    Auterinen, I.; Salmenhaara, S.E.J. . Author

    2004-01-01

    The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

  18. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  19. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  20. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  1. Natural Circulation Capability Assessments for a Small-medium Reactor

    International Nuclear Information System (INIS)

    Choi, Sun Do

    2010-02-01

    Small-medium reactors have been highly evaluated to have more safe characteristics than those of large reactors. In addition, it could be used for a variety of purposes, such as small-scale power production in mountainous of island area, seawater desalination, regional heating system. For a higher safety, studies about a way of using natural circulation have being conducted around world. CAREM(Argentina), AST- 500(Russia), and NHR-200(china) etc. According to this tendency, REX- 10(Regional Energy rX-10) is designed in Korea for regional heating and small-scale power production. To investigate the thermal-hydraulic behavior of REX-10, we designed Rex-10 Test Facility (RTF), simulating REX-10, by using the scaling law. The scaling ratios of length, volume and power were set with 1/1, 1/50 and 1/50, respectively. The diameter and total length of RTF are 40 cm and approximately 6 m, respectively. The facility is composed of various components, which are a core in the bottom part, a heat exchanger in the middle part, a pressurizer and hot legs in the upper part, and chillers outside the facility. The test instrumentation is also designed to measure temperatures, flow rates, pressures, and pressure drop. The experiment parameters were adopted based on the 1-dimensional approach. There are a variety of parameters which influence natural circulation behavior such as heater power, overall flow resistance parameter, the distance between the center of the heat exchanger and the core. As the experimental geometries are fixed, it is found that the most important parameter is the heater power under the experimental conditions. In addition, to evaluate the effect of heater power, some experiments were conducted at varying heater power condition (from 70 kW to 170 kW) under constant primary pressure (2.0 MPa) and secondary flow rate (4.5 liter per minute). As the results of the experiments, the temperature and flow rate increase with increasing heater power. The flow rate is

  2. Nuclear material control at IEA-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    1988-01-01

    The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

  3. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    1999-12-31

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

  4. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits

  5. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

  6. Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

    International Nuclear Information System (INIS)

    Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

    2010-01-01

    In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

  7. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

  8. 1DB, a one-dimensional diffusion code for nuclear reactor analysis

    International Nuclear Information System (INIS)

    Little, W.W. Jr.

    1991-09-01

    1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k eff and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor formnd to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs

  9. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  10. A comparative study on inhibition of total astragalus saponins and astragaloside IV on TNFR1-mediated signaling pathways in arterial endothelial cells.

    Directory of Open Access Journals (Sweden)

    Qin-she Liu

    Full Text Available Both total astragalus saponins (AST and it's main component astragaloside IV (ASIV have been used in China as cardiovascular protective medicines. However, the anti-inflammatory activities that are beneficial for cardiovascular health have never been compared directly and the molecular mechanisms remain unresolved. This study was conducted to compare the inhibitory effects of these drugs on TNFα-induced cell responses, related signaling pathways, and the underlying mechanisms in mouse arterial endothelial cells.Real-time qRT-PCR was performed to determine the expression of cell adhesion molecule (CAM genes. Immunofluorescent staining was used to detect the nuclear translocation of transcription factor NF-κB-p65. Western Blot analysis was used to identify TNFα-induced NF-κB-p65 phosphorylation, IκBα degradation, and caspase-3 cleavage. Cell surface proteins were isolated and TNFα receptor-1(TNFR1 expression was determined. The results suggest that both AST and ASIV attenuate TNFα-induced up-regulation of CAMs mRNA and upstream nuclear translocation and phosphorylation of NF-κB-p65. However, TNFR1-mediated IκBα degradation, cleavage of caspase-3 and apoptosis were inhibited only by AST. These differences in the actions of AST and ASIV could be explained by the presence of other components in AST, such as ASII and ASIII, which also had an inhibitory effect on TNFR1-induced IκBα degradation. Moreover, AST, but not ASIV, was able to reduce TNFR1 protein level on the cell surface. Furthermore, mechanistic investigation demonstrated that TNFR1-mediated IκBα degradation was reversed by the use of TAPI-0, an inhibitor of TNFα converting enzyme (TACE, suggesting the involvement of TACE in the modulation of surface TNFR1 level by AST.ASIV was not a better inhibitor than AST, at least on the inhibition of TNFα-induced inflammatory responses and TNFR1-mediated signaling pathways in AECs. The inhibitory effect of AST was caused by the

  11. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  12. RA reactor operation and maintenance in 1989, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Sanovic, V.

    1989-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  13. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  14. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  15. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  16. Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.

    1988-01-01

    Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

  17. Core calculations for the upgrading of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

    1998-01-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  18. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  19. Dose measurements in controlled area and laboratory of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Alvarenga, Frederico Ladeia

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers. (author)

  20. Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

    2014-08-15

    IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

  1. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  2. Operating reactors licensing actions summary. Volume 5, Number 1

    International Nuclear Information System (INIS)

    1985-03-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

  3. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    International Nuclear Information System (INIS)

    Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

    2008-01-01

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

  4. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

    2008-10-15

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

  5. DETERMINAÇÃO DOS VALORES MÉDIOS DAS ENZIMAS AST, DHL, gGT E FAS NO SORO DE EQUINOS SADIOS EM SANTA MARIA, RS

    Directory of Open Access Journals (Sweden)

    Sônia Terezinha dos Anjos Lopes

    1993-12-01

    Full Text Available Foram usados 50 eqüinos sadios provenientes do Batalhão de Polícia Montada da Brigada Militar em Santa Maria, RS, sendo 43 machos e 7 fêmeas com idade variadas a partir de 3 anos. Foram colhidos 10ml de sangue da jugular para determinação dos valores da atividade sérica das enzimas aspartato-aminotransferase (AST, desidrogenase lática (DHL, gama-glutamiltransferase (gGT e fosfatase alcalina sérica (FAS. Os resultados encontrados para AST foi de 101 - 190U/I com média de 130UI; DHL foi de 100 - 421 U/l com média de 182U/I; gGT foi de 2 - 27U/I com média de 6.5U/I e FAS foi de 103 - 335U/I com média de 190U/I. A partir de outubro/1992 estes valores passaram a ser referência no laboratório de Patologia Clínica do Hospital de Clinicas Veterinárias da Universidade Federal de Santa Maria.

  6. Source term determination from subcritical multiplication measurements at Koral-1 reactor

    International Nuclear Information System (INIS)

    Blazquez, J.B.; Barrado, J.M.

    1978-01-01

    By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

  7. Studies in fusion reactor technology. Final report, September 1, 1974--August 31, 1977

    International Nuclear Information System (INIS)

    Axtmann, R.C.; Perkins, H.K.

    1977-08-01

    Two independent measurements of hydrogen permeation through stainless steel at driving pressures in the range from 10 -6 to 1 Pa indicate that most extant predictions of tritium permeation through fusion reactors are probably overestimated grossly. A comprehensive analysis demonstrates that, given available structural materials, the prospects are negligible for the economic production of synthetic fuels via radiolytic reactions in fusion reactor systems

  8. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  9. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    International Nuclear Information System (INIS)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

    2015-01-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  10. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  11. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  12. AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

    International Nuclear Information System (INIS)

    Hansen, W.

    2003-01-01

    Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de

  13. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Balogun, G.I.; Jonah, S.A.; Umar, I.M.

    2005-01-01

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  14. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1995-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy's Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period

  15. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

  16. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  17. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  18. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  19. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  20. Calculation of the main neutron parameters of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Ojima, Mario Katsuhiko

    1977-01-01

    The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

  1. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  2. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  3. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  4. Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

    International Nuclear Information System (INIS)

    Kulcinski, G.L.; Maynard, C.W.

    1984-04-01

    Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors

  5. PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

    1990-06-01

    Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs

  6. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    Miller, R.L.

    1997-01-01

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  7. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

    1965-08-15

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.

  8. Neutron radiography in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

    1988-08-01

    Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

  9. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

  10. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  11. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

    International Nuclear Information System (INIS)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

    2014-10-01

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  12. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  13. Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

    International Nuclear Information System (INIS)

    Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

    2013-06-01

    (Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

  14. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  15. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

  16. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  17. Measurements of reactivity of reactor G1

    International Nuclear Information System (INIS)

    Bernot, J.; Koechlin, J.C.; Portes, L.; Teste du Bailler, A.

    1957-01-01

    The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [fr

  18. Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

    International Nuclear Information System (INIS)

    Larry B. Wimmer

    2001-01-01

    The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000)

  19. Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

    International Nuclear Information System (INIS)

    Moreira, Joao Manoel Losada; Frajndlich, Roberto

    1995-01-01

    This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

  20. Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

    2017-08-30

    The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO3-xPbTiO3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.

  1. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  2. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

    CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...

  3. Reactor of the XXI century

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.; Solov'ev, Yu.A.

    2001-01-01

    The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

  4. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  5. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  6. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  7. Extension of cycle 8 of Angra-1 reactor, optimization of electric power generation reduction

    International Nuclear Information System (INIS)

    Miranda, Anselmo Ferreira; Moreira, Francisco Jose; Valladares, Gastao Lommez

    2000-01-01

    The main objective of extending fuel cycle length of Angra-1 reactor, is in fact of that each normal refueling are changed about 40 fuel elements of the reactor core. Considering that these elements do not return for the reactor core, this procedure has became possible a more gain of energy of these elements. The extension consists in, after power generation corresponding to a cycle burnup of 13700 MWD/TMU or 363.3 days, to use the reactivity gain by reduction of power and temperature of primary system for power generation in a low energy patamar

  8. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  9. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  10. The Osiris reactor. Descriptive report - Volume 1 - text

    International Nuclear Information System (INIS)

    1969-05-01

    Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation [fr

  11. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  12. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  13. Monitoring of primary circuit and reactor of NPP A-1

    International Nuclear Information System (INIS)

    Prazska, M.; Majersky, M.; Rezbarik, J.; Sekely, S.; Vozarik, P.; Walthery, R.; Stuller, P.

    2005-01-01

    Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m 2 . It follows that the total gamma contamination is of the order of 10 14 to 10 15 Bq and total alpha contamination 10 11 to 10 13 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)

  14. Characterisation of reactor control rod drives. Specification 1-6. Reaktorstellstabantriebe. Typenblaetter 1-6

    Energy Technology Data Exchange (ETDEWEB)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN).

  15. The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

    International Nuclear Information System (INIS)

    Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2004-01-01

    The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

  16. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  17. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  18. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  19. Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

    International Nuclear Information System (INIS)

    Duque O, J.; Velez A, L.H.

    1987-01-01

    Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

  20. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1973-01-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  1. Characterisation of reactor control rod drives. Specification 1-6

    International Nuclear Information System (INIS)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de

  2. Developing maintainability in controlled thermonuclear reactors. Progress report, October 1, 1977--April 30, 1978

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1977-05-01

    During the period 1 October 1977 through 30 April 1978 the study has completed work on Task 6, Candidate Reference Systems. Four candidate reference systems have been defined. These are based on the conceptual designs of the UWMAK-III, the General Atomic Company Demonstration Power Reactor, the Oak Ridge National Laboratory Cassette defined in the Demonstration Power Study and the Culham laboratory Mark II Reactors. These reactor concepts are normalized to 3000 MW/sub th/ and near minimum cost of electricity. In addition, designs of four major subsystems have been selected and defined for application to these reactors. These include a primary coolant system, primary and secondary vacuum zone systems, the neutral beam injection system and the magnetic field system. These magnet systems are unique to each reactor. The cases for which maintenance plans are being developed in Task 7 have been selected to allow evaluation of design features, particularly the vacuum wall locations, and the impacts of unscheduled and contact maintenance of subsystems on the cost of electricity

  3. Ferritin levels, inflammatory biomarkers, and mortality in peripheral arterial disease: a substudy of the Iron (Fe) and Atherosclerosis Study (FeAST) Trial.

    Science.gov (United States)

    Depalma, Ralph G; Hayes, Virginia W; Chow, Bruce K; Shamayeva, Galina; May, Patricia E; Zacharski, Leo R

    2010-06-01

    This study delineated correlations between ferritin, inflammatory biomarkers, and mortality in a cohort of 100 cancer-free patients with peripheral arterial disease (PAD) participating in the Veterans Affairs (VA) Cooperative Study #410, the Iron (Fe) and Atherosclerosis Study (FeAST). FeAST, a prospective, randomized, single-blind clinical trial, tested the hypothesis that reduction of iron stores using phlebotomy would influence clinical outcomes in 1227 PAD patients randomized to iron reduction or control groups. The effects of statin administration were also examined in the Sierra Nevada Health Care (SNHC) cohort by measuring serum ferritin levels at entry and during the 6-year study period. No difference was documented between treatment groups in all-cause mortality and secondary outcomes of death plus nonfatal myocardial infarction and stroke. Iron reduction in the main study caused a significant age-related improvement in cardiovascular disease outcomes, new cancer diagnoses, and cancer-specific death. Tumor necrosis factor (TNF)-alpha, TNF-alpha receptors 1 and 2, interleukin (IL)-2, IL-6, IL-10, and high-sensitivity C reactive protein (hs-CRP) were measured at entry and at 6-month intervals for 6 years. Average levels of ferritin and lipids at entry and at the end of the study were compared. The clinical course and ferritin levels of 23 participants who died during the study were reviewed. At entry, mean age of entry was 67 +/- 9 years for the SNHCS cohort, comparable to FeAST and clinical and laboratory parameters were equivalent in substudy participants randomized to iron reduction (n = 51) or control (n = 49). At baseline, 53 participants on statins had slightly lower mean entry-level ferritin values (114.06 ng/mL; 95% confidence interval [CI] 93.43-134.69) vs the 47 off statins (127.62 ng/mL; 95% CI, 103.21-152.02). Longitudinal analysis of follow-up data, after adjusting for the phlebotomy treatment effect, showed that statin use was associated with

  4. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

    1999-01-01

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  5. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  6. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  7. Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

    2009-01-01

    The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

  8. Review of NHR activities in the Russian Federation

    International Nuclear Information System (INIS)

    Malamud, V.A.; Kurachenkov, A.V.; Kusmartsev, E.V.

    1997-01-01

    NHR development activities in the ex-USSR were initiated in the 1970s mainly due to a growing deficiency of organic fuels needed for heating large cities in the European part of the country. Construction of two pilot nuclear district heating plants with AST-500 NHRs was started in the early 1980s, and by 1989 the first unit in Gorky NDHP was nearly 90% completed. Current activity in this field is concentrated on upgrading the AST-500 design and on the development on this basis of a whole series of heating-only and co-generation reactor plants of unit power ranging from 30 to 600 MW. A brief description of the AST-500 reference NHR design features is given, as well as of the R and D activities that have been carried out for the design decisions and safety validation. (author). 12 refs, 1 tab

  9. Review of NHR activities in the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Malamud, V A; Kurachenkov, A V; Kusmartsev, E V [OKBM, Nizhny Novgorod (Russian Federation)

    1997-09-01

    NHR development activities in the ex-USSR were initiated in the 1970s mainly due to a growing deficiency of organic fuels needed for heating large cities in the European part of the country. Construction of two pilot nuclear district heating plants with AST-500 NHRs was started in the early 1980s, and by 1989 the first unit in Gorky NDHP was nearly 90% completed. Current activity in this field is concentrated on upgrading the AST-500 design and on the development on this basis of a whole series of heating-only and co-generation reactor plants of unit power ranging from 30 to 600 MW. A brief description of the AST-500 reference NHR design features is given, as well as of the R and D activities that have been carried out for the design decisions and safety validation. (author). 12 refs, 1 tab.

  10. Probabilistic risk analysis of Angra-1 reactor

    International Nuclear Information System (INIS)

    Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.

    1986-01-01

    The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt

  11. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

    1987-01-01

    The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

  12. Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

    International Nuclear Information System (INIS)

    Ashoub, N.; Saleh, H.

    1995-01-01

    The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

  13. Integral tightness measurements at the Paks-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taubner, R.; Techy, Z. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

    1983-01-01

    The containment system experiments of the Paks-1 nuclear reactor are described. The integrated tightness measurements of the hermetic system were completed in 1982. The principles and methods and the evaluation of the results of the measurements are discussed. Some features of the filtration characteristics are demonstrated using relative values and a method enabling the description of the physical contents of the characteristics by flow technical functions is outlined.

  14. Activity of Catalase (CAT, ALT and AST in Different Organs of Swiss Albino Mice Treated with Lead Acetate, Vitamin C and Magnesium-L-Threonate

    Directory of Open Access Journals (Sweden)

    Ilir Nazmi Mazreku

    2017-11-01

    Full Text Available Introduction: Lead is a natural element with toxic properties and is widespread in the environment. Lead toxicity is associated with generation of reactive oxygen and nitrogen species and consumption of antioxidants elements (vitamin E and C, glutathione, thioredoxin and lipoic acid, melatonin, carotenoids and natural flavonoids in the cell, and unbalancing oxidantsantioxidants levels. Aim: To evaluate the effects of different chemical combinations (lead acetate, Vitamin C and Magnesium-L-threonate on antioxidant enzyme activity (catalase-CAT of liver, kidney, spleen, pancreas and brain, and serum transaminases [Serum Alanine Transaminase (ALT and Serum Aspartate Transaminase (AST]. Materials and Methods: Experimental animals (49 male Mus musculus-swiss albino mice were separated into five different groups. The first group was used as a control, hence the other four groups were treated with sub-lethal doses (90 mg/kg of lead acetate (group 2, lead acetate (90 mg/kg and Vitamin C dose 40mg/kg (group 3, lead acetate (90 mg/kg and Magnesium-Lthreonate dose 100 mg/kg (group 4 and only with MagnesiumL-threonate dose 100 mg/kg (group 5, during the treatment period (40 days. Blood samples were taken from the facial vein and used for transaminase analysis. Organ tissue was collected after euthanizing anaesthetized animals with neck dislocation technique. Results: The results showed that lead acetate treatment has caused significant elevation in the activity of AST (group 2 and 3 and ALT (group 3. Also, CAT activity was significantly (p<0.05 increased in groups treated with lead acetate (liver, pancreas, kidney and brain but not in spleen. Treatment of lead intoxicated groups with Vitamin C and Magnesium L-threonate increased significantly CAT activity in brain. Conclusion: Lead effects by interacting with different molecular systems and increasing enzyme activity (CAT, ALT and AST. Effects on CAT activity of Magnesium-L-threonate and Vitamin C treatment

  15. Molten-salt reactor strategies viewed from fuel conservation effect, (1)

    International Nuclear Information System (INIS)

    Furuhashi, Akira

    1976-01-01

    Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

  16. Security devices and experiment facilities at ENEA TRIGA RC-1 reactor

    International Nuclear Information System (INIS)

    Bianchi, P.; Festinesi, A.; Santoro, E.; Tardani, G.; Magli, M.; Reis, G.

    1990-01-01

    RC-1 TRIGA operating exercise staff has produced some auxiliary security devices. These are the neutron source automatic handling device, irradiated samples rabbit connection rotating rack, and auxiliary equipment for transferring hot fuel elements. The reactor electronic control instrumentation system includes various instrumentation channels, the operating capability of which must be verified by the licensee as per Italian regulations. In order to obtain automatic and repeatable operations, TEMAV designed and constructed a remotely-driven source transfer device, based on requirements, performance specifications and technical data supplied by ENEA-TIB. The pneumatic irradiating system for short lived materials allows extraction of radiated samples in a time no longer than 4 seconds. To optimize the system, both as to operability and health protection, a specific rotating rack for the connection of irradiated samples with pneumatic transfer (RABBIT) was produced. To permit 1 MW hot fuel element storage in pits it is necessary to remove hot 100 KW fuel elements and transfer them to a re-treatment plant. Feasibility studies showed the impossibility of using heavy trucks inside the reactor hall. To avoid problems trucks are left outside the reactor hall and only the PEGASO container is removed with a special device that runs on rails. Movement from Rail truck is assured by an electromotor driving pull device and security cable

  17. Qualification process of dispersion fuels in the IEAR1 research reactor

    International Nuclear Information System (INIS)

    Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

  18. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  19. Membrane-aerated biofilm reactor for the removal of 1,2-dichloroethane by Pseudomonas sp strain DCA1

    NARCIS (Netherlands)

    Hage, J.C.; Houten, R.T.; Tramper, J.; Hartmans, S.

    2004-01-01

    A membrane-aerated biofilm reactor (MBR) with a biofilm of Pseudomonas sp. strain DCA1 was studied for the removal of 1,2-dichloroethane (DCA) from water. A hydrophobic membrane was used to create a barrier between the liquid and the gas phase. Inoculation of the MBR with cells of strain DCA1 grown

  20. TRAC-BD1: transient reactor analysis code for boiling-water systems

    International Nuclear Information System (INIS)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

  1. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1)

    International Nuclear Information System (INIS)

    Forlerer, Elena; Palacios, Tulio A.

    1998-01-01

    After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation

  2. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    Mathet, Eric; Castello, Gerard

    2012-01-01

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  3. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included.

  4. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included

  5. University Reactor Sharing Program. Period covered: September 1, 1981-August 31, 1982

    International Nuclear Information System (INIS)

    Hajek, B.K.; Myser, R.D.; Miller, D.W.

    1982-12-01

    During the period from September 1, 1981 to August 31, 1982, the Ohio State University Nuclear Reactor Laboratory participated in the Reactor Sharing Program by providing services to eight colleges and universities. A laboratory on Neutron Activation Analysis was developed for students in the program. A summary of services provided and a copy of the laboratory procedure are attached. Services provided in the last funded period were in three major areas. These were neutron activation analysis, nuclear engineering labs, and introductions to nuclear research. One group also performed radiation surveys and produced isotopes for calibration of their own analytical equipment

  6. Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, A. P. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Matos, J. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-02-28

    The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations of the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.

  7. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  8. Upgrade of Instrumentation for Purdue Reactor PUR-1

    International Nuclear Information System (INIS)

    Revankar, S.T.; Merritt, E.; Bean, R.

    2000-01-01

    The major objective of this program was to upgrade and replace instruments and equipment that significantly improve the performance, control and operational capability of the Purdue University nuclear reactor (PUR-1). Under this major objective two projects on instrument upgrade were implemented. The first one was to convert the vacuum tube control and safety amplifiers (CSA) to solid state electronics, and the other was to upgrade the electrical and electronic shielding. This report is the annual report and gives the efforts and progress achieved on these two projects from July 1999 to June 2000

  9. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    Blom, K.H.; Krull, W.

    1997-01-01

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

  10. Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

    Energy Technology Data Exchange (ETDEWEB)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-07-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  11. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  12. Thermal-hydraulic design of the 200 MW NHR

    Energy Technology Data Exchange (ETDEWEB)

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The main problems regarding the AST-500 NHR thermal-hydraulics are considered. Basic thermal data of the reactor plant are given and peculiarities of coolant parameters at natural convection in the primary circuit are discussed. The in-reactor instrumentation system is briefly describes, as well as the results of natural-convective flow characteristics investigations using reactor test models. (author). 4 refs, 5 figs.

  13. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  14. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  15. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  16. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Rizo, O; Alvarez, I; Herrera, E; Lima, L; Tores, J [Secretaria Ejecutiva para Asuntos Nucleares, Holguin (Cuba). Delegacion Territorial; Manso, M V [Centro de Isotopos, La Habana (Cuba); Lopez, M C [Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico); Ixquiac, M [Universidad de San Carlos de Guatemala, Guatemala City (Guatemala)

    1997-12-31

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K{sub o} neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott`s formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented.

  17. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

    2011-01-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  18. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  19. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  20. Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    Ricardo, Carla Pereira; Mesquita, Amir Zacarias

    2007-01-01

    During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

  1. The different generation of nuclear reactors from Generation-1 to Generation-4

    International Nuclear Information System (INIS)

    Cognet, G.

    2010-01-01

    In this work author deals with the history of the development of nuclear reactors from Generation-1 to Generation-4. The fuel cycle and radioactive waste management as well as major accidents are presented, too.

  2. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

  3. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  4. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  5. Shiga toxin 1 induces on lipopolysaccharide-treated astrocytes the release of tumor necrosis factor-alpha that alter brain-like endothelium integrity.

    Directory of Open Access Journals (Sweden)

    Verónica I Landoni

    Full Text Available The hemolytic uremic syndrome (HUS is characterized by hemolytic anemia, thrombocytopenia and renal dysfunction. The typical form of HUS is generally associated with infections by Gram-negative Shiga toxin (Stx-producing Escherichia coli (STEC. Endothelial dysfunction induced by Stx is central, but bacterial lipopolysaccharide (LPS and neutrophils (PMN contribute to the pathophysiology. Although renal failure is characteristic of this syndrome, neurological complications occur in severe cases and is usually associated with death. Impaired blood-brain barrier (BBB is associated with damage to cerebral endothelial cells (ECs that comprise the BBB. Astrocytes (ASTs are inflammatory cells in the brain and determine the BBB function. ASTs are in close proximity to ECs, hence the study of the effects of Stx1 and LPS on ASTs, and the influence of their response on ECs is essential. We have previously demonstrated that Stx1 and LPS induced activation of rat ASTs and the release of inflammatory factors such as TNF-α, nitric oxide and chemokines. Here, we demonstrate that rat ASTs-derived factors alter permeability of ECs with brain properties (HUVECd; suggesting that functional properties of BBB could also be affected. Additionally, these factors activate HUVECd and render them into a proagregant state promoting PMN and platelets adhesion. Moreover, these effects were dependent on ASTs secreted-TNF-α. Stx1 and LPS-induced ASTs response could influence brain ECs integrity and BBB function once Stx and factors associated to the STEC infection reach the brain parenchyma and therefore contribute to the development of the neuropathology observed in HUS.

  6. Shiga Toxin 1 Induces on Lipopolysaccharide-Treated Astrocytes the Release of Tumor Necrosis Factor-alpha that Alter Brain-Like Endothelium Integrity

    Science.gov (United States)

    Landoni, Verónica I.; Schierloh, Pablo; de Campos Nebel, Marcelo; Fernández, Gabriela C.; Calatayud, Cecilia; Lapponi, María J.; Isturiz, Martín A.

    2012-01-01

    The hemolytic uremic syndrome (HUS) is characterized by hemolytic anemia, thrombocytopenia and renal dysfunction. The typical form of HUS is generally associated with infections by Gram-negative Shiga toxin (Stx)-producing Escherichia coli (STEC). Endothelial dysfunction induced by Stx is central, but bacterial lipopolysaccharide (LPS) and neutrophils (PMN) contribute to the pathophysiology. Although renal failure is characteristic of this syndrome, neurological complications occur in severe cases and is usually associated with death. Impaired blood-brain barrier (BBB) is associated with damage to cerebral endothelial cells (ECs) that comprise the BBB. Astrocytes (ASTs) are inflammatory cells in the brain and determine the BBB function. ASTs are in close proximity to ECs, hence the study of the effects of Stx1 and LPS on ASTs, and the influence of their response on ECs is essential. We have previously demonstrated that Stx1 and LPS induced activation of rat ASTs and the release of inflammatory factors such as TNF-α, nitric oxide and chemokines. Here, we demonstrate that rat ASTs-derived factors alter permeability of ECs with brain properties (HUVECd); suggesting that functional properties of BBB could also be affected. Additionally, these factors activate HUVECd and render them into a proagregant state promoting PMN and platelets adhesion. Moreover, these effects were dependent on ASTs secreted-TNF-α. Stx1 and LPS-induced ASTs response could influence brain ECs integrity and BBB function once Stx and factors associated to the STEC infection reach the brain parenchyma and therefore contribute to the development of the neuropathology observed in HUS. PMID:22479186

  7. Feasibility study of application of Prompt Gamma Neutron Activation Analysis (PGNAA) method in TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Guerra, Bruno Teixeira

    2016-01-01

    The TRIGA Mark I IPR-R1 research reactor is located at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. The reactor operates at 100 kW but the core configuration allows the increasing of the power up to 250 kW. It has been applied research, training and radioisotopes production. The establishment of the Prompt Gamma Neutron Activation Analysis (PGNAA) method at the TRIGA IPR-R1 reactor will significantly increase the types of matrices analysed as well as the number of chemical elements. Additionally it will complement the neutron activation analysis. This work presents a proposed design of a PGNAA facility to be installed at the TRIGA IPR-R1. The proposed design is based on a tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. Thus, the aim of this study is to verify the feasibility to establish the PGNAA method in IPR-R1 through theoretical study applying the Monte Carlo code. The feasibility of establishing the PGAA method at the IPR-R1 installations was evaluated through of the calculations of neutron flux, radioactive capture reaction rates and detection limits for some isotopes. According to the obtained results, it can be concluded that is possible to establish the PGAA method at the IPR-R1 reactor, even with some restrictions in its theoretical design calculated by MCNP. (author)

  8. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  9. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  10. Application of Cherenkov light observation to reactor measurements (1). Estimation of reactor power from Cherenkov light intensity

    International Nuclear Information System (INIS)

    Yamamoto, Keiichi; Takeuchi, Tomoaki; Kimura, Nobuaki; Ohtsuka, Noriaki; Tsuchiya, Kunihiko; Sano, Tadafumi; Nakajima, Ken; Homma, Ryohei; Kosuge, Fumiaki

    2015-01-01

    Development of the reactor measurement system was started to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. The calculation and the observation were performed to obtain the quantity of the Cherenkov light caused by the gamma and beta rays emitted from the fuels in the core of Kyoto University Research Reactor. The results indicate that the real-time reactor power can be estimated from the brightness of the Cherenkov light observed by a CCD camera. This method can also work for the estimation of the burn-up of spent fuels at commercial reactors. Since the observed brightness value of the Cherenkov light was influenced by the camera position, the optical observation method should be improved to achieve high accuracy observation. (author)

  11. Measurements and calculations of reactivity for the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.; Maiorino, J.R.; Yamaguchi, M.

    1988-01-01

    This work shows a measurement of reactivity parameters, such as integral and diferential control rod worth, local void coefficient, and moderator temperature coefficient for the research reactor IEA-R1. The measured values were compared with those calculated through HAMMER-CITATION codes, having shown good agreement. (author) [pt

  12. Accident management strategies for VVER-1000 reactors. Part 1: text

    International Nuclear Information System (INIS)

    Sdouz, G.; Sonneck, G.; Pachole, M.

    1994-10-01

    This report describes the effect of different accident management strategies on the onset, development and end of the core-concrete-interaction as well as on the containment integrity for a TMLB'-type sequence in a Pressurized Water Reactor of the type VVER- 1000. Using the computer code MARCH3 the following strategies were investigated: (1) One or more Spray and LP ECC Systems available with and without coolers after 10 hours (2) Inclusion of the reactor pressure vessel testing facility room to the cavity (3) Containment venting (4) External water supply and (5) Different electric power restoration times. The results show that some of these accident management measures will maintain the containment integrity and reduce the source term drastically, others will reduce the source term rate. For some measures final conclusions can only be given after complete source term calculations have been performed. (authors)

  13. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  14. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  15. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  16. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  17. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  18. Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

  19. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

    2013-01-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  20. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  1. Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

    International Nuclear Information System (INIS)

    Hutton, P.H.; Pappas, R.A.; Friesel, M.A.

    1985-02-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs

  2. Small propulsion reactor design based on particle bed reactor concept

    International Nuclear Information System (INIS)

    Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

    1989-01-01

    In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

  3. Space reactor electric systems: system integration studies, Phase 1 report

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-01-01

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

  4. Thermal performance of Egypt's research reactor core (ET-RR-1)

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, A.

    1986-01-01

    The steady state thermal performance of the ET-RR-1 core system is theoretically investigated by different models describing the heat flux and the coolant mass flow rate. The magnitude of the heat generated by a fuel element depends upon its position in the core. Normal and uniform distributions for heat flux and coolant mass flow rate are considered. The clad and coolant temperatures at different core positions are evaluated and compared with the experimental measurements at different operating conditions. The results indicated large discrepancy between the predicted and the experimental results. Therefore, the previous models and the experimental results are evaluated in order to develop the best model that describes the thermal performance of the ET-RR-1 core. The adapted model gives 99.5% significant confidence limit. The effect of increasing the heat flux or decreasing the mass flow rate by 20% from its maximum recommended operating condition is tested and discussed. Also, the thermal behaviour towards increasing the reactor power more than its maximum operating condition is discussed. The present work could also be used in extending the investigation to other PWR reactor operating conditions

  5. Calibration of new I and C at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, Martin; Jurickova, Monika

    2011-01-01

    The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

  6. Neutronics Design of Helical Type DEMO Reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Sagara, A.; Goto, T.; Yanagi, N.; Masuzaki, S.; Tamura, H.; Miyazawa, J.; Muroga, T., E-mail: teru@nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: Neutronics design study has been performed in a newly started conceptual design activity for a helical type DEMO reactor FFHR-d1. Features of the FFHR-d1 design are enlargement of the basic configurations of reactor components and extrapolation of plasma parameters from those of the helical type plasma experimental machine Large Helical Device (LHD) to achieve the highest feasibility. From the neutronics point of view, a blanket space of FFHR-d1 is severely limited at the inboard of the torus. This is due to the core plasma position shifting to the inboard side under the confinement condition extrapolated from LHD. The first step of the neutronics investigation using the MCNP code has been performed with a simple torus model simulating thin inboard blanket space. A Flibe+Be/Ferritic steel breeding blanket showed preferable performances for both tritium breeding and shielding, and has been adapted as a reference blanket system for FFHR-d1. The investigations indicate that a combination of a 15 cm thick breeding blanket, 55 cm thick WC+B4C shield, i.e., the blanket space of 70 cm, could suppress the fast neutron flux and nuclear heating in the helical coils to the design targets for the neutron wall loading of 1.5 MW/m{sup 2}. Since the outboard side can provide a large space for a 60 cm thick breeding blanket, a fully-covered tritium breeding ratio (TBR) of 1.31 has been obtained in the simple torus model. The neutronics design study has proceeded to the second step using a 3-D helical reactor model. The most important issue in the 3-D neutronics design is a compatibility with the helical divertor design. To achieve a higher TBR and shielding performance, the core plasma has to be covered by the breeding blanket layers as possible. However, the dimensions of the blanket layers are limited by magnetic field lines connecting an edge of the core plasma and divertor pumping ports. After repeating modification of the blanket configuration, the global TBR of 1

  7. Evaluation of applicability of alternative source terms to operating nuclear power plants in Korea

    International Nuclear Information System (INIS)

    Lim, S. N.; Park, Y. S.; Nam, K. M.; Song, D. B.; Bae, Y. J.; Lee, Y. J.; Jung, C. Y.

    2002-01-01

    In 1995 and 2000, NRC issued NUREG-1465 and Regulatory Guide 1.183 with respect to Alternative Source Terms(AST) replacing the existing source terms of TID-14844 and Regulatory Guide 1.4, 1.25, and 1.77 for radiological Design Basis Accidents(DBA) analysis. In 1990, ICRP published ICRP Pub. 60 which represents new recommendations on dose criteria and concepts. In Korea, alternative source terms were used for evaluation of effective doses for design basis accidents of Advanced Power Reactor(APR1400) using the computer program developed by an overseas company. Recently, DBADOSE, new computer program for DBA analysis incorporating AST and effective dose concept was developed by KHNP and KOPEC, and reanalysis applying AST to operating nuclear power plants, Kori units 3 and 4 in Korea using DBADOSE has been performed. As the results of this analysis, it was concluded that some conservative variables or operation procedures of operating plants could be mitigated or simplified by virtue of increased safety margin and consequently, economical and operational benefits ensue. In this paper, methodologies and results of Kori 3 and 4 DBA reanalysis and sensitivity analysis for mitigation of main design variables are introduced

  8. HECTR [Hydrogen Event Containment Transient Response] Version 1.5N: A modification of HECTR Version 1.5 for application to N Reactor

    International Nuclear Information System (INIS)

    Camp, A.L.; Dingman, S.E.

    1987-05-01

    This report describes HECTR Version 1.5N, which is a special version of HECTR developed specifically for application to the N Reactor. HECTR is a fast-running, lumped-parameter containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. Version 1.5N is a modification of Version 1.5 and includes changes to the spray actuation logic, and models for steam vents, vacuum breakers, and building cross-vents. Thus, all of the key features of the N Reactor confinement can be modeled. HECTR is designed for flexibility and provides for user control of many important parameters, if built-in correlations and default values are not desired

  9. What occurred in the reactors

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko

    2013-01-01

    Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

  10. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  11. The role of SASSYS-1 in LMR [Liquid Metal Reactor] safety analysis

    International Nuclear Information System (INIS)

    Dunn, F.E.; Wei, T.Y.C.

    1988-01-01

    The SASSYS-1 liquid metal reactor systems analysis computer code is currently being used as the principal tool for analysis of reactor plant transients in LMR development projects. These include the IFR and EBR-II Projects at Argonne National Laboratory, the FFTF project at Westinghouse-Hanford, the PRISM project at General Electric, the SAFR project at Rockwell International, and the LSPB project at EPRI. The SASSYS-1 code features a multiple-channel thermal-hydraulics core representation coupled with a point kinetics neutronics model with reactivity feedback, all combined with detailed one-dimensional thermal-hydraulic models of the primary and intermediate heat transport systems, including pipes, pumps, plena, valves, heat exchangers and steam generators. In addition, SASSYS-1 contains detailed models for active and passive shutdown and emergency heat rejection systems and a generalized plant control system model. With these models, SASSYS-1 provides the capability to analyze a wide range of transients, including normal operational transients, shutdown heat removal transients, and anticipated transients without scram events. 26 refs., 16 figs

  12. Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Souza, Rose Mary Gomes do Prado

    2009-01-01

    Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R1 TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculate as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. (author))

  13. Experimental investigations on the transient behaviour of nuclear heat plants with natural convection

    International Nuclear Information System (INIS)

    Adam, E.; Sydow, J.; Wolff, J.

    1988-01-01

    Apart from the theoretical approach, practical experiments concerning the transient behaviour of the primary loop of reactors with natural coolant convection are necessary in order to evaluate the safety systems of reactors providing heat for industrial and communal consumers. The article presents experiments concerning the transient behaviour of the experimental plant DANTON, which models the reactor AST-500, and gives a preview of further research. (orig.) [de

  14. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  15. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  16. First fuel re-load of Angra-1 reactor - Inspection and hearing plan

    International Nuclear Information System (INIS)

    Pollis, W.; Alvarenga, M.A.B.; Meldonian, N.L.; Paiva, R.L.C. de; Pollis, R.

    1985-01-01

    The plan of inspection and hearing of the first fuel reload of Angra-1 nuclear reactor is detailed. It consists in five steps: receiving and storage of the fuel; reload preparation; activities during; post-reload activities, and preliminary activities. (M.I.)

  17. Multipurpose RTOF Fourier diffractometer at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Maayouf, R.M.A.; Tiitta, A.T.

    1993-09-01

    The present work represents a further study of the basic RTOF Fourier multipurpose diffractometer, to start with, at the ET-RR-1 reactor. The functions of the suggested arrangement are thoroughly discussed and the possibilities if its expansion are also assessed. The flexibility of the arrangement allows its further expansion both for stress measurement at 90 deg. scattering angle with two detector banks at opposite sides of the incident beam and for operation in the transmission diffraction mode. (orig.). (19 refs., 10 figs., 1 tab.)

  18. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  19. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  20. Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

    International Nuclear Information System (INIS)

    Auterinen, Iiro; Salmenhaara, Seppo

    2008-01-01

    FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

  1. Neutron Spectrum Parameters In Inner Irradiation Channel Of The Nigeria Research Reactor-1 (NIRR-1) For Use In Absolute And KO-NAA Methods

    International Nuclear Information System (INIS)

    Jonah, S.A; Balogun, G.I; Mayaki, M.C.

    2004-01-01

    In Nigeria, the first Nuclear Reactor achieved critically on February 03, 2004 at about 11:35 GMT and has been commissioned or training and research. It is a Miniature Neutron Source Reactor (MNSR), code-named Nigeria Research Reactor-1 (NIRR-1). NIRR-1 has a tan-in-pool structural configuration and a nominal thermal power rating of 30 Kw. With a built-in clean old core excess reactivity of 3.77 mk determined during the on-site zero and critically experimental, the reactor can operate for a n.cm-2 .s-1 in the inner irradiation channels). Under these conditions, the reactor can operate with the same fuel loading for over ten years with a burn-up of <1%. A detailed description of operating characteristics for NIRR-1, measured during the on-site zero-power and criticality experiments has been given elsewhere. In order to extend its utilization to include absolute and ko-NAA methods, the neutron spectrum parameters in the irradiation channels: power and critically experiments has been given elsewhere. In order to extend it's the irradiation channels: thermal-to-epithermal flux ration, F; and epithermal flux shape factor, a in both the inner and outer irradiation channels must be determined experimentally. In this work, we have developed and experimental procedure for monitoring the neutron spectrum parameters in an inner irradiation channel based on irradiation and gamma-ray counting of detector foils via (n,y), (n,p) and (n,a) dosimetry reactions. Results obtained indicate that a thermal neutron flux of (5.14+-0.02) x 1011 n/c m2.s determined by foil activation method in the inner irradiation channel, B2, at a power level of 15.5 kw corresponds to the flux indicators on the control console and the micro-computer control system respectively. Other parameters of the neutron spectrum determined for inner irradiation channel B2, are: a -0.0502+0.003; 18.92+-0.14; F = 3.87=0.23. The method was validated through the comparison of our result with published neutron spectrum

  2. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA, Deo 1 - Pogon, odrzavanje i eksploatacija reaktora u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Milosevic, M; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1981-12-15

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  3. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  4. Maintenance of reactor recirculation pumps [Paper No.: II-1

    International Nuclear Information System (INIS)

    Ansari, M.A.; Bhat, K.P.

    1981-01-01

    At Tarapur Atomic Power Station (TAPS), two reactor recirculation pumps are provided, one each for the two reactor units. The performance of pumps has been uniformly good; however, leakage through the cartridge type, two stage, mechanical seals which are installed on these pumps was encountered on few occasions. The paper describes the leakage problems, identification of certain design deficiencies and rectification carried out at TAPS for overcoming these problems. (author)

  5. Study on operational aspect of natural circulation HLMC reactor (1)

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

    2000-08-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

  6. Design and development of weld inspection manipulator for reactor pressure vessel of TAPS-1

    International Nuclear Information System (INIS)

    Chatterjee, H.; Singh, J.P.; Ranjon, R.; Kulkarni, M.P.; Patel, R.J.

    2013-01-01

    The reactor pressure vessel (RPV) of TAPS-1 BWR contains six longitudinal and four circumferential welds. Periodical in-service inspection of these weld joints has been a regulatory issue pending for long. In the 22 nd refuelling outage in July 2012 the inspection of L1-1, L1-2 longitudinal welds as well as their junctions with C1 circumferential weld were proposed to be done using ultrasonic technique. Approaching these welds from OD side of the RPV is a difficult and tedious task. Therefore it was decided to examine these welds from ID side of the RPV by filling the cavity with water and approaching the RPV from top. No technology was locally available to take the probes at a depth of 10-12 m under water. NPCIL approached RTD, BARC to develop an underwater manipulator to accomplish this task. RTD took up this work as a challenge and came out with the design of manipulator. The weld inspection manipulator (WIM) was fabricated on a war foot basis, tested and successfully implemented in the reactor for the first time in TAPS history. The entire activity was completed in three months time. This article gives the details of design, manufacturing, performance testing, qualification trials and implementation of WIM in the reactor. Ultrasonic testing techniques were developed by QAD, BARC which are not covered in this article. (author)

  7. Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

    International Nuclear Information System (INIS)

    Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

    2005-01-01

    A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

  8. Neutronics analysis of Nigerian Research Reactor-1

    International Nuclear Information System (INIS)

    Azande, T.S.; Balogun, G.I.

    2010-01-01

    Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.

  9. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M and calculated (C results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE/C ratios of 1.10 for both neutron (E >1.0 MeV flux and iron atom displacement rate.

  10. Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

    International Nuclear Information System (INIS)

    Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

    2014-01-01

    Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

  11. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  12. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  13. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  14. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  15. Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2015-01-01

    The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

  16. Quality assurance measures at the Geesthacht research reactor FRG-1

    International Nuclear Information System (INIS)

    Voss, J.; Krull, W.; Schmidt, K.

    1995-01-01

    The major part of the quality system for the FRG is already in practical use; other parts require extensive preparations and therefore transition periods of different lengths before they can be introduced. GKSS is aware that beside the other upgrading and refurbishment activities, these duality assurance measures will be very important in ensuring the operation of the FRG-1 research reactor over the coming 15 years or more. (orig.)

  17. Aspects of the Iea-R1 research reactor seismic evaluation

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    1996-01-01

    Codes and standards for the seismic evaluation of the research reactor IEA-R1 are presented. An approach to define the design basis earthquake based on the local seismic map and on simplified analysis methods is proposed. The site seismic evaluation indicates that the design earthquake intensity is IV MM. Therefore, according to the used codes and standards, no buildings, systems, and components seismic analysis are required. (author)

  18. Modernization of the CDTN IPR-R1 TRIGA reactor instrumentation and control

    International Nuclear Information System (INIS)

    Mesquita, A.Z.; Costa, A.C.L.; Souza, R.M.G.P.

    2009-01-01

    The control system of the IPR-R1 was changed in 1995. Although since the year's 80 was generalized the use of microprocessor technology and video monitors for visual interface, in the IPR-R1 control room it was used analogical system by relay-based logic, and were maintained the mechanical strip chart recorders (ink-pen drive) to measure, monitor and store the operational parameters. It was maintained the measure and the control of, practically, the same variables of the original system, although the reactor power already have been upgraded to 100 kW and began the studies to increase it to 250 kW, which is the current core configuration. For 250 kW operations the fuel heat transfer becomes important and new parameters should be used as safety operational limits. A state-of-the-art instrumentation and control system using microprocessor technology is proposed to replace the present analogical systems. The new system can eliminates most manual data logging, provides automatic or manual reactor operation modes, provides complete real-time operator display, replays historical operating data on monitor or printer, eliminates spare parts replacement problems and meets all applicable international standards as NRC and IEE specifications. This paper describes the research project in process in CDTN that has as objective the modernization of the IPR-R1 TRIGA reactor instrumentation and control of the operational variables. The project also will improve the accomplishment of neutronic and thermal-hydraulic experiments, foreseen in the CDTN research program. (author)

  19. Physical modelling of the composting environment: A review. Part 1: Reactor systems

    International Nuclear Information System (INIS)

    Mason, I.G.; Milke, M.W.

    2005-01-01

    In this paper, laboratory- and pilot-scale reactors used for investigation of the composting process are described and their characteristics and application reviewed. Reactor types were categorised by the present authors as fixed-temperature, self-heating, controlled temperature difference and controlled heat flux, depending upon the means of management of heat flux through vessel walls. The review indicated that fixed-temperature reactors have significant applications in studying reaction rates and other phenomena, but may self-heat to higher temperatures during the process. Self-heating laboratory-scale reactors, although inexpensive and uncomplicated, were shown to typically suffer from disproportionately large losses through the walls, even with substantial insulation present. At pilot scale, however, even moderately insulated self-heating reactors are able to reproduce wall losses similar to those reported for full-scale systems, and a simple technique for estimation of insulation requirements for self-heating reactors is presented. In contrast, controlled temperature difference and controlled heat flux laboratory reactors can provide spatial temperature differentials similar to those in full-scale systems, and can simulate full-scale wall losses. Surface area to volume ratios, a significant factor in terms of heat loss through vessel walls, were estimated by the present authors at 5.0-88.0 m 2 /m 3 for experimental composting reactors and 0.4-3.8 m 2 /m 3 for full-scale systems. Non-thermodynamic factors such as compression, sidewall airflow effects, channelling and mixing may affect simulation performance and are discussed. Further work to investigate wall effects in composting reactors, to obtain more data on horizontal temperature profiles and rates of biological heat production, to incorporate compressive effects into experimental reactors and to investigate experimental systems employing natural ventilation is suggested

  20. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  1. Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

  2. Study of the $B^0 \\rightarrow K^{\\ast 0}e^+ e^-$ decay with the LHCb detector and development of a novel concept of PID detector: the Focusing DIRC

    CERN Document Server

    AUTHOR|(CDS)2088184; Arnaud, Nicolas

    Flavour-changing neutral current processes of the type $b \\to s\\gamma$ are forbidden at the tree level in the Standard Model (SM) and occur at leading order through radiative loop diagrams. Therefore, they are sensitive to new physics (NP), which may contribute with competing diagrams. Furthermore, the chirality of the weak interaction in the SM implies that the photon emitted has left-handed polarisation. However, a whole class of NP theories do not share this SM feature and may manifest unambiguously as a right-handed contribution to the polarisation. This thesis presents the first study of the ${b}{s\\gamma}$ photon polarisation through an angular analysis of the $B^0 \\rightarrow K^{\\ast 0}e^+ e^-$ channel. Even though $B^0 \\rightarrow K^{\\ast 0}e^+ e^-$ is not a radiative $b\\to s$ transition, the contribution from a virtual photon coupling to the lepton pair dominates in the low-$q^2$ region. Furthermore, the channel with electrons rather than muons allows to better isolate the virtual photon contribution ...

  3. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  4. Flow-induced vibration test of an advanced water reactor model. Pt. 1. Turbulence-induced forcing function

    International Nuclear Information System (INIS)

    Au-Yang, M.K.; Brenneman, B.; Raj, D.

    1995-01-01

    A 1:9 scale model of a proposed advanced water reactor was tested for flow-induced vibration. The main objectives of this test were: (1) to derive an empirical equation for the turbulence forcing function which can be applied to the full-sized prototype; (2) to study the effect of viscosity on the turbulence; (3) to verify the ''superposition'' assumption widely used in dynamic analysis of weakly coupled fluid-shell systems; and (4) to measure the shell responses to verify methods and computer programs used in the flow-induced vibration analysis of the prototype. This paper describes objectives (1), (2), and (3); objective (4) will be discussed in a companion paper.The turbulence-induced fluctuating pressure was measured at 49 locations over the surface of a thick-walled, non-responsive scale model of the reactor vessel/core support cylinders. An empirical equation relating the fluctuating pressure, the frequency, and the distance from the inlet nozzle center line was derived to fit the test data. This equation involves only non-dimensional, fluid mechanical parameters that are postulated to represent the full-sized, geometrically similar prototype. While this postulate cannot be verified until similar measurements are taken on the full-sized unit, a similar approach using a 1:6 scale model of a commercial pressurized water reactor was verified in the mid-1970s by field measurements on the full-sized reactor. (orig.)

  5. Problems of nuclear reactor safety. Vol. 1; Problemy bezopasnosti yaderno-ehnergeticheskikh ustanovok. Tom 1

    Energy Technology Data Exchange (ETDEWEB)

    Shal` nov, A V [Moskovskij Inzhenerno-Fizicheskij Inst., Moscow (Russian Federation)

    1996-12-31

    Proceedings of the 9. Topical Meeting `Problems of nuclear reactor safety` are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP`s are analyzed.

  6. The future of the IPR-R1 TRIGA MARK I reactor after 48 years operation

    International Nuclear Information System (INIS)

    Maretti, Fausto Junior; Sette Camara, Luiz Otavio I.; Oliveira, Paulo Fernando

    2008-01-01

    The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center/ Brazilian Committion for Nuclear Energy (CDTN/CNEN), originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it operated for isotope production for different uses, being later used in wide scale for another purposes as analyses for activation with neutrons and training of nuclear power plants operators. Dozens of degree theses were also developed with the use of the reactor. Along the years, several improvements were introduced in the reactor and its auxiliary systems, with the purpose to provide better use of the facilities and with the objective to increase the safety in the operation. The reactor is ready right now to operate at 250 kW, and for sure the nuclear applications programmed will be improved. The Operation Manual and the Safety Analysis report were already modified, as well as the Emergency Plan and the relative procedures to the same. After the tests at the end of 2008, the reactor will already be operating in the new power. This work presents a description of the several accomplishments of the last years and comments about the possibility of new uses for the reactor in the several areas of nuclear applications and some of the experiments and tests results during the upgrading program. (authors)

  7. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

    International Nuclear Information System (INIS)

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

    A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

  8. New digital control and power protection system of VR 1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Juoeickova, M.

    2005-01-01

    The contribution describes the new VR-1 training reactor control and power protection system at the Czech Technical University in Prague. The control system provides safety and control functions, calculates average values of the important variables and sends data and system status to the human-machine interface. The upgraded control system is based on a high quality industrial PC. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The software was developed according to requirements in MS Visual C. The independent power protection system is a component of the reactor safety (protection) system with high quality and reliability requirements. The digital system is redundant; each channel evaluates the reactor power and the velocity of power changes and provides safety functions. The digital part of the channel is multiprocessor-based. The software was developed with respect to nuclear standards. The software design was coded in the C language regarding the NRC restrictions. Configuration management, verification and validation accompanied the software development. Both systems were thoroughly tested. Firstly, the non active tests were carried out. During these tests, the active core of the reactor was subcritical; the input signals were generated from HPIB and VXI controlled instruments to simulate different operational and safety events. The software for instruments control and tests evaluation utilized Agilent VEE development system. After the successful non active checking, the active tests followed. (author)

  9. Modelling of the RA-1 reactor using a Monte Carlo code

    International Nuclear Information System (INIS)

    Quinteiro, Guillermo F.; Calabrese, Carlos R.

    2000-01-01

    It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

  10. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  11. 75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on...

    Science.gov (United States)

    2010-06-24

    ... Standard Review Plan Section 13.6.1, Revision 1 on Physical Security--Combined License and Operating...), Section 13.6.1 on ``Physical Security--Combined License and Operating Reactors,'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100350158). The Office of Nuclear Security and...

  12. Evaluation on radioactive waste disposal amount of Kori Unit 1 reactor vessel considering cutting and packaging methods

    International Nuclear Information System (INIS)

    Choi, Yu Jong; Lee, Seong Cheol; Kim, Chang Lak

    2016-01-01

    Decommissioning of nuclear power plants has become a big issue in South Korea as some of the nuclear power plants in operation including Kori unit 1 and Wolsung unit 1 are getting old. Recently, Wolsung unit 1 received permission to continue operation while Kori unit 1 will shut down permanently in June 2017. With the consideration of segmentation method and disposal containers, this paper evaluated final disposal amount of radioactive waste generated from decommissioning of the reactor pressure vessel in Kori unit 1 which will be decommissioned as the first in South Korea. The evaluation results indicated that the final disposal amount from the top and bottom heads of the reactor pressure vessel with hemisphere shape decreased as they were cut in smaller more effectively than the cylindrical part of the reactor pressure vessel. It was also investigated that 200 L and 320 L radioactive waste disposal containers used in Kyung-Ju disposal facility had low payload efficiency because of loading weight limitation

  13. Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

    International Nuclear Information System (INIS)

    Castiblanco, L.A.; Sarta, J.A.

    1997-01-01

    The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

  14. The SPHINX reactor for engineering tests

    International Nuclear Information System (INIS)

    Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

    1990-01-01

    A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

  15. Application of safety checklist to the analysis of the IEA-R1 reactor water retreatment system

    International Nuclear Information System (INIS)

    Sauer, Maria Eugenia Lago Jacques; Sara Neto, Antonio Jorge; Lima, Toni Carlos Caboclo de; Ribeiro, Maria Alice Morato

    2005-01-01

    In 1999, the management of the IEA-R1 Research Reactor (pool type - 5 MWth), located at IPEN/CNEN-SP, started the evaluation of the Reactor Pool Water Retreatment System to identify operational aspects, which could compromise the operators safety. The purpose was to identify and propose enhancements to the system which would be installed to substitute for the existing one. This process was conducted through a qualitative study of the system in operation. This study was carried out by a team composed of specialists in reactor operation, systems maintenance and radiological protection, and one safety analyst. The study consisted, basically, in local inspections to verify the physical and operational conditions of each equipment / component as well as aspects related to maintenance activities of the system. The process control and the operator procedures associated with the retreatment of the reactor pool water were also reviewed. The methodology adopted to develop the study was based in process hazard analysis technique named Safety Checklist. This paper presents a summary of this study and the main results obtained. Some operational and safety problems identified, the prevention and/or correction means to avoid them, and the recommendations and suggestions that have been implemented to the new design of the IEA-R1 Reactor Water Retreatment System, whose installation was concluded in 2003, are also presented. (author)

  16. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    D. Kokkinos

    2005-01-01

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  17. Upgrade of VR-1 training reactor I and C

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Chab, V.

    2003-01-01

    The contribution describes the upgrade of the VR-1 training reactor I and C (Instrumentation and Control). The reactor was put into operation in the 1990, and its I and C seems to be obsolete now. The new I and C utilises the same digital technology as the old one. The upgrade has been done gradually during holidays in order not to disturb the reactor utilisation during teaching and training. The first stage consisted in the human-machine interface and the control room upgrade in 2001. A new operator's desk, displays, indicators and buttons were installed. Completely new software and communication interface to the present I and C were developed. During the second stage in 2002, new control rod drivers and safety circuits were installed. The rod motors were replaced and necessary mechanical changes on the control rod mechanism, induced by the utilisation of the new motor, were done. The new safety circuits utilise high quality relays with forced contacts to guarantee high reliability of their operation. The third stage, the control system upgrade is being carried out now. The new control system is based on an industrial PC mounted in a 19 inch crate. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. A large amount of work has been devoted to the software requirements to specify all dependencies, modes and permitted actions, safety measures, etc. The Department took an active part in the setting of software requirements and later in verification and validation of the software and the whole control system. Finally, a new protection system consisting of power measuring and power protection channels will be installed in 2004 or 2005. (author)

  18. Data acquisition and signal processing system for IPR R1 TRIGA-Mark I nuclear research reactor of CDTN

    International Nuclear Information System (INIS)

    Mesquita, A.Z.; Maretti, F. Jr.; Rezende, H.C.; Tambourgi, E.B.

    2004-01-01

    The TRIGA IPR-R1 Nuclear Research Reactor, located at the Nuclear Technology Development Center (CDTN/CNEN) in Belo Horizonte, Brazil, is being operated since 44 years ago. The main operational parameters were monitored by analog recorders and counters located in the reactor control console. The reactor operators registered the most important operational parameters and data in the reactor logbook. This process is quite useful, but it can involve some human errors. It is also impossible for the operators to take notes of all variables involving the process mainly during fast power transients in some operations. A PC-based data acquisition was developed for the reactor that allows online monitoring, through graphic interfaces, and shows operational parameters evolution to the operators. Some parameters that were not measured, like the power and the coolant flow rate at the primary loop, are monitored now in the computer video monitor. The developed system allows measuring out all parameters in a frequency up to 1 kHz. These data is also recorded in text files available for consults and analysis. (author)

  19. International nuclear reactor hazard study. Design and operational features, and hazards of commercial nuclear power reactors in the world. Vol. 1

    International Nuclear Information System (INIS)

    Anderson, R.; Benecke, J.; Donderer, R.; Etemad, S.; Hirsch, H.; Kjellstroem, B.; Large, J.; Scheer, J.; Schneider, M.; Schumacher, H.; Schumacher, O.; Scott, M.; Takagi, J.; Thompson, G.; Torrie, R.; Ziggel, H.

    1986-09-01

    Project management and scientific coordination was performed by Gruppe Oekologie Hannover. The experts from Canada, the Federal Republic of Germany, France, Great Britain, Sweden, USA, and Japan represented an annulated experience of work on nuclear safety. The panel members analyzed the reactor types of their respective countries. The draft reports on the individual reactor types were discussed. Also included are brief descriptions of the different reactor types with simple drawing, evolution of reactor types and a brief report on nuclear power plants in the world. (DG)

  20. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  1. Determination of the theoretical and experimental zero-power frequency response of Ghana Research Reactor-1

    International Nuclear Information System (INIS)

    Intsiful, J.D.K.; Akaho, E.H.K.; Tetteh, G.K.

    1997-12-01

    The frequency response measurements of a reactor at low power help in determining the kinetic parameters of a reactor and ultimately in investigating its stability with respect to small perturbations in reactivity. In this report, we present the results of the zero-power frequency response measurements of GHARR-1 by rod method and its analytical analogue. The comparison in calculated and measured values is reasonably good in the frequency range used (author)

  2. UPGRADE OF INSTRUMENTATION FOR PURDUE REACTOR PUR-1, PHASE 3

    International Nuclear Information System (INIS)

    Revankar, S. T.

    2004-01-01

    The major objective of this program is to upgrade and replace instruments and equipment that significantly improve the performance, control and operational capability of the Purdue University nuclear reactor (PUR-1). Under this major objective one project on design and installation of interface cards for channel four detector was considered. This report is the final report and gives the efforts and progress achieved on these projects from August 2002 to July 2004

  3. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  4. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

  5. Modernization of turbine control system and reactor control system in Almaraz 1 and 2

    International Nuclear Information System (INIS)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-01-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  6. ISO-9001: An approach to accreditation for an MTR facility: SAFARI-1 research reactor

    International Nuclear Information System (INIS)

    Piani, C.S.B.; Du Bruyn, J.F.B.

    2000-01-01

    The SAFARI-1 Research Reactor obtained ISO-9001 accreditation via the South African Bureau of Standards in September 1998. In view of the commercial applications of the reactor, the value of acquisition of the accreditation was considered against the cost of implementation of the Quality System. The criteria identified in the ISO-9001 standard were appraised and a superstructure derived for management of the generation and implementation of a suitable Quality Management System (QMS) for the fairly unique application of a nuclear research reactor. A Quality Policy was established, which formed the basis of the QMS against which the various requirements and/or standards were identified. In addition, since it was considered advantageous to incorporate the management controls of Conventional and Radiological Safety as well as Plant Maintenance and Environmental Management (ISO 14001), these aspects were included in the QMS. (author)

  7. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  8. Thermodynamic properties of helium in the range from 20 to 15000C and 1 to 100 bar. Reactor core design of high-temperature gas-cooled reactors. Pt. 1

    International Nuclear Information System (INIS)

    Kipke, H.E.; Stoehr, A.; Banerjea, A.; Hammeke, K.; Huepping, N.

    1978-12-01

    The following report presents in tabular form the safety standard of the nuclear safety standard commission (KTA) on reactor core design of high-temperature gas-cooled reactors. Part 1: Calculation of thermodynamic properties of helium The basis of the present work is the data and formulae given by H. Petersen for the calculation of density, specific heat, thermal conductivity and dynamic viscosity of helium together with the formula for their standard deviations in the range of temperature and pressure stated above. The relations for specific enthalpy and specific entropy have been derived from density and specific heat, whereby specific heat is assumed constant over the given range of temperature and pressure. The latter section of this report contains tables of thermodynamic properties of helium calculated from the equations stated earlier in this paper. (orig.) [de

  9. Reactor dynamics experiment of N.S. Mutsu using pseudo random signal. 1

    International Nuclear Information System (INIS)

    Hayashi, Koji; Nabeshima, Kunihiko; Shinohara, Yoshikuni; Shimazaki, Junya; Inoue, Kimihiko; Ochiai, Masaaki.

    1993-10-01

    In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, reactor noise experiments using pseudo random binary sequences (PRBS) have been planned, and a preliminary experiment was performed on March 4, 1991 in the first experimental navigation with the aim of checking the experimental procedures and conditions. The experiments using both reactivity and load disturbances were performed at 70 % of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. From the results obtained, we confirmed that (1) the procedures and experimental conditions determined prior to the experiment were suitable for performing the PRBS experiments, (2) when the PRBS disturbances were applied, the plant state remained quite stable, and (3) the quality of the measured data is adequate for the purpose of dynamics analysis. This paper summarizes the planning and preparation of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

  10. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

    International Nuclear Information System (INIS)

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

    This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

  11. Thermal neutron flux measurements in the rotary specimen rack of the IPR-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G. do Prado; Rodrigues, Rogério R.; Souza, Luiz Claudio A., E-mail: souzarm@cdtn.br, E-mail: rrr@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The thermal neutron flux in the rotary specimen rack of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil, has been measured by the neutron activation method, using bare and cadmium covered gold foils. Those foils were irradiated in the rotary specimen rack with the reactor at 100 kW. The reactor core configuration has 63 fuel elements, composed of 59 original aluminum-clad elements and 4 stainless steel-clad fuel elements. The gamma activities of the foils were measured using Ge spectrometer. The perturbations of the thermal neutron flux caused by the introduction of an absorbing foil into the medium were considered in order to obtain accurate determination of the flux. The thermal neutron flux obtained was 7.4 x 10{sup 11} n.cm{sup -2}.s{sup -1}. (author)

  12. Ethanol production from wet-exploded wheat straw hydrolysate by thermophilic anaerobic bacterium Thermoanaerobacter BG1L1 in a continuous immobilized reactor

    DEFF Research Database (Denmark)

    Georgieva, Tania I.; Mikkelsen, Marie Just; Ahring, Birgitte Kiær

    2008-01-01

    was not detoxified, ethanol yield in a range of 0.39-0.42 g/g was obtained. Overall, sugar efficiency to ethanol was 68-76%. The reactor was operated continuously for approximately 143 days, and no contamination was seen without the use of any agent for preventing bacterial infections. The tested microorganism has......Thermophilic ethanol fermentation of wet-exploded wheat straw hydrolysate was investigated in a continuous immobilized reactor system. The experiments were carried out in a lab-scale fluidized bed reactor (FBR) at 70C. Undetoxified wheat straw hydrolysate was used (3-12% dry matter), corresponding...... to sugar mixtures of glucose and xylose ranging from 12 to 41 g/l. The organism, thermophilic anaerobic bacterium Thermoanaerobacter BG1L1, exhibited significant resistance to high levels of acetic acid (up to 10 g/l) and other metabolic inhibitors present in the hydrolysate. Although the hydrolysate...

  13. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

    2014-01-01

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia

  14. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Mazleha Maskin; Phongsakorn, P.T.; Tonny, A.L.; Fedrick, C.M.B.; Faizal Mohamed; Mohamad Fauzi Saad; Ahmad Razali Ismail; Mohamad Puad Haji Abu

    2013-01-01

    Full-text: As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia. (author)

  15. Development of a reactor thermalhydraulic experiment databank(SORTED1)

    International Nuclear Information System (INIS)

    Bang, Young Seck; Kim, Eun Kyoung; Kim, Hho Jung; Lee, Sang Yong

    1994-01-01

    The recent trend in thermalhydraulic safety analysis of nuclear power plant shows the best-estimate and probabilistic approaches, therefore, the verification of the best-estimate code based on the applicable experiment data has been required. The present study focused on developing a simple databank, SORTED1, to be effectively used for code verification. The development of SORTED1 includes a data collection from the various sources including ENCOUNTER, which is the reactor safety data bank of U.S. Nuclear Regulatory Commission, a reorganization of collected resources suitable for requirements of SORTED1 database management system (DBMS), and a development of a simple DBMS. The SORTED1 is designed in Unix environment with graphic user interface to improve a user convenience and has a capability to provide the test related information. The currently registered data in SORTED1 cover 759 thermalhydraulic tests including LOFT, Semiscale, etc

  16. The advanced MAPLE reactor concept

    International Nuclear Information System (INIS)

    Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

    1989-01-01

    High-flux neutron sources are continuing to be of interest both in Canada and internationally to support materials testing for advanced power reactors, new developments in extracted-neutron-beam applications, and commercial production of selected radioisotopes. The advanced MAPLE reactor concept has been developed to meet these needs. The advanced MAPLE reactor is a new tank-type D 2 O reactor that uses rodded low-enrichment uranium fuel in a compact annular core to generate peak thermal-neutron fluxes of 1 x 10 19 n·s -1 in a central irradiation rig with a thermal power output of 50 MW. Capital and incremental development costs are minimized by using MAPLE reactor technology to the greatest extent practicable

  17. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

    1988-05-01

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  18. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  19. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  20. ENSI's view on technical safety for the long term operation of reactors 1 and 2 in the Beznau nuclear power plant

    International Nuclear Information System (INIS)

    2010-11-01

    The reactors 1 and 2 of the Beznau nuclear power plant (KKB) are operated since about 40 years. For an operation beyond the design period of 40 years the Swiss Federal Nuclear Safety Inspectorate (ENSI) demands the evidence to be brought that the design limits of the safety relevant components will not be reached during the extended operation period. In 2008 the license holder of KKB delivered the requested documentation on material ageing on the basis of deterministic as well as probabilistic safety analyses and concluded that both reactors can be safely operated beyond 40 years. Thanks to continuous additional outfits, both reactors are in good condition from the point of view of technical safety. With a view to the extension of operation beyond 40 years, KKB already applied the necessary measures regarding technics, finances and personnel in order to keep the present technical level. Since 1991 KKB has analysed and checked components that are difficult to replace. From the evidence presented, ENSI concluded that both reactors are able to be operated up to 60 years long, however with two restrictions for reactor 1 because there the material used for the reactor pressure vessel (RPV) suffered more neutron brittleness than in reactor 2. In addition, reactor 1 is much more affected by ageing phenomena than reactor 2, but, according to neutron fluence calculations, the limiting criteria will not be reached even after 60 years of operation. Some corrosion damages were noted at the lower part of the RPV due to water containing boron acid; they are more pronounced in reactor 1 than in reactor 2. Even though the calculations done by KKB are very conservative, they show that also in the long term the operation limiting criteria about the mechanical resistance of the RPV are never reached. ENSI concludes that the safety design of both KKB reactors ensures safe control of the design basis accidents. Both reactors were continuously fitted with new equipment. With the planed