Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

International Nuclear Information System (INIS)

Lajtman, S.; Feretic, D.; Debrecin, N.

1989-01-01

This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

Validation of the Serpent 2-DYNSUB code sequence using the Special Power Excursion Reactor Test III (SPERT III)

International Nuclear Information System (INIS)

Knebel, Miriam; Mercatali, Luigi; Sanchez, Victor; Stieglitz, Robert; Macian-Juan, Rafael

2016-01-01

Highlights: • Full few-group cross section tables created by Monte Carlo lattice code Serpent 2. • Serpent 2 group constant methodology verified for HFP static and transient cases. • Serpent 2-DYNSUB tool chainvalidated using SPERT III REA experiments. • Serpent 2-DYNSUB tool chain suitable to model RIAs in PWRs. - Abstract: The Special Power Excursion Reactor Test III (SPERT III) is studied using the Serpent 2-DYNSUB code sequence in order to validate it for modeling reactivity insertion accidents (RIA) in PWRs. The SPERT III E-core was a thermal research reactor constructed to analyze reactor dynamics. Its configuration resembles a commercial PWR on terms of fuel type, choice of moderator, coolant flow and system pressure. The initial conditions of the rod ejection accident experiments (REA) performed cover cold startup, hot startup, hot standby and operating power scenarios. Eight of these experiments were analyzed in detail. Firstly, multi-dimensional nodal diffusion cross section tables were created for the three-dimensional reactor simulator DYNSUB employing the Monte Carlo neutron transport code Serpent 2. In a second step, DYNSUB stationary simulations were compared to Monte Carlo reference three-dimensional full scale solutions obtained with Serpent 2 (cold startup conditions) and Serpent 2/SUBCHANFLOW (operating power conditions) with a good agreement being observed. The latter tool is an internal coupling of Serpent 2 and the sub-channel thermal-hydraulics code SUBCHANFLOW. Finally, DYNSUB was utilized to study the eight selected transient experiments. Results were found to match measurements well. As the selected experiments cover much of the possible transient (delayed super-critical, prompt super-critical and super-prompt critical excursion) and initial conditions (cold and hot as well as zero, little and full power reactor states) one expects in commercial PWRs, the obtained results give confidence that the Serpent 2-DYNSUB tool chain is

Fabrication and characterization of SPR chips with the modified bovine serum albumin

Science.gov (United States)

Chen, Xing; Zhang, Lu-lu; Cui, Da-fu

2016-03-01

A facile surface plasmon resonance (SPR) chip is developed for small molecule determination and analysis. The SPR chip was prepared based on a self assembling principle, in which the modified bovine serum albumin (BSA) was directly self-assembled onto the bare gold surface. The surface morphology of the chip with the modified BSA was investigated by atomic force microscopy (AFM) and its optical properties were characterized. The surface binding capacity of the bare facile SPR chip with a uniform morphology is 8 times of that of the bare control SPR chip. Based on the experiments of immune reaction between cortisol antibody and cortisol derivative, the sensitivity of the facile SPR chip with the modified BSA is much higher than that of the control SPR chip with the un-modified BSA. The facile SPR chip has been successfully used to detect small molecules. The lowest detection limit is 5 ng/mL with a linear range of 5—100 ng/mL for cortisol analysis. The novel facile SPR chip can also be applied to detect other small molecules.

SMORN-III benchmark test on reactor noise analysis methods

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Hirota, Jitsuya

1984-02-01

A computational benchmark test was performed in conjunction with the Third Specialists Meeting on Reactor Noise (SMORN-III) which was held in Tokyo, Japan in October 1981. This report summarizes the results of the test as well as the works made for preparation of the test. (author)

SPR 2015. Abstracts

International Nuclear Information System (INIS)

2015-01-01

The volume contains the abstracts of the SPR (society for pediatric radiology) 2015 meeting covering the following issues: fetal imaging, muscoskeletal imaging, cardiac imaging, chest imaging, oncologic imaging, tools for process improvement, child abuse, contrast enhanced ultrasound, image gently - update of radiation dose recording/reporting/monitoring - meaningful or useless meaning?, pediatric thoracic imaging, ALARA.

SPR 2015. Abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-04-01

The volume contains the abstracts of the SPR (society for pediatric radiology) 2015 meeting covering the following issues: fetal imaging, muscoskeletal imaging, cardiac imaging, chest imaging, oncologic imaging, tools for process improvement, child abuse, contrast enhanced ultrasound, image gently - update of radiation dose recording/reporting/monitoring - meaningful or useless meaning?, pediatric thoracic imaging, ALARA.

Nano Sensing and Energy Conversion Using Surface Plasmon Resonance (SPR

Directory of Open Access Journals (Sweden)

Iltai (Isaac Kim

2015-07-01

Full Text Available Nanophotonic technique has been attracting much attention in applications of nano-bio-chemical sensing and energy conversion of solar energy harvesting and enhanced energy transfer. One approach for nano-bio-chemical sensing is surface plasmon resonance (SPR imaging, which can detect the material properties, such as density, ion concentration, temperature, and effective refractive index in high sensitivity, label-free, and real-time under ambient conditions. Recent study shows that SPR can successfully detect the concentration variation of nanofluids during evaporation-induced self-assembly process. Spoof surface plasmon resonance based on multilayer metallo-dielectric hyperbolic metamaterials demonstrate SPR dispersion control, which can be combined with SPR imaging, to characterize high refractive index materials because of its exotic optical properties. Furthermore, nano-biophotonics could enable innovative energy conversion such as the increase of absorption and emission efficiency and the perfect absorption. Localized SPR using metal nanoparticles show highly enhanced absorption in solar energy harvesting. Three-dimensional hyperbolic metamaterial cavity nanostructure shows enhanced spontaneous emission. Recently ultrathin film perfect absorber is demonstrated with the film thickness is as low as ~1/50th of the operating wavelength using epsilon-near-zero (ENZ phenomena at the wavelength close to SPR. It is expected to provide a breakthrough in sensing and energy conversion applications using the exotic optical properties based on the nanophotonic technique.

SPR imaging combined with cyclic voltammetry for the detection of neural activity

Directory of Open Access Journals (Sweden)

Hui Li

2014-03-01

Full Text Available Surface plasmon resonance (SPR detects changes in refractive index at a metal-dielectric interface. In this study, SPR imaging (SPRi combined with cyclic voltammetry (CV was applied to detect neural activity in isolated bullfrog sciatic nerves. The neural activities induced by chemical and electrical stimulation led to an SPR response, and the activities were recorded in real time. The activities of different parts of the sciatic nerve were recorded and compared. The results demonstrated that SPR imaging combined with CV is a powerful tool for the investigation of neural activity.

Tritium handling, breeding, and containment in two conceptual fusion reactor designs: UWMAK-II and UWMAK-III

International Nuclear Information System (INIS)

Clemmer, R.G.; Larsen, E.M.

1976-01-01

Tritium is an essential component of near-term controlled thermonuclear reactor systems. Since tritium is not likely to be available on a large scale at a modest cost, fusion reactor designs must incorporate blanket systems which will be capable of breeding tritium. Because of the radiological activity and capability of assimilation into living tissues, tritium release to the environment must be strictly controlled. The University of Wisconsin has completed three conceptual designs of fusion reactors, UWMAK-I, UWMAK-II, and UWMAK-III. This report discusses the tritium systems for UWMAK-II, a reactor design with a helium cooled solid breeder blanket, and UWMAK-III, a reactor design with a high-temperature liquid breeder blanket. Tritium systems for fueling and recycling, breeding and recovery, and plant containment and control are discussed. (Auth.)

Enhanced gas sensing performance of TiO2 functionalized magneto-optical SPR sensors

OpenAIRE

Manera, Maria Grazia; Montagna, G.; Ferreiro-Vila, Elías; González-García, Lola; Sánchez-Valencia, J.R.; González-Elipe, Agustín R.; Cebollada, Alfonso; García-Martín, José Miguel; García-Martín, Antonio; Armelles Reig, Gaspar; Rella, Roberto

2011-01-01

Porous TiO2 thin films deposited by glancing angle deposition are used as sensing layers to monitor their sensing capabilities towards Volatile Organic Compounds both in a standard Surface Plasmon Resonance (SPR) sensor and in Magneto-Optical Surface Plasmon Resonance (MO-SPR) configuration in order to compare their sensing performances. Here our results on the enhanced sensing capability of these TiO2 functionalized MO-SPR sensors with Au/Co/Au transducers with respect to traditional SPR gas...

SPR 2014. Abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-05-15

The proceedings of the SPR 2014 meeting include abstracts on the following topics: Body imaging techniques: practical advice for clinic work; thoracic imaging: focus on the lungs; gastrointestinal imaging: focus on the pancreas and bowel; genitourinary imaging: focus on gonadal radiology; muscoskeletal imaging; focus on oncology; child abuse and nor child abuse: focus on radiography; impact of NMR and CT imaging on management of CHD; education and communication: art and practice in pediatric radiology.

SPR 2014. Abstracts

International Nuclear Information System (INIS)

2014-01-01

The proceedings of the SPR 2014 meeting include abstracts on the following topics: Body imaging techniques: practical advice for clinic work; thoracic imaging: focus on the lungs; gastrointestinal imaging: focus on the pancreas and bowel; genitourinary imaging: focus on gonadal radiology; muscoskeletal imaging; focus on oncology; child abuse and nor child abuse: focus on radiography; impact of NMR and CT imaging on management of CHD; education and communication: art and practice in pediatric radiology.

Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

International Nuclear Information System (INIS)

Wang, Shisheng; Rineiski, Andrei; Maschek, Werner; Ignatiev, Victor

2006-01-01

Recent developments extending the capabilities of the SIMMER-III code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmuter. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed. (authors)

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor; Normativa aplicable y desarrollo de experimentos de vigilancia de aproximacion a criticidad en el reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J L; Aguilar H, F; Rivero G, T; Sainz M, E [Instituto nacional de Investigaciones Nucleares, Departamento de Automatizacion, A.P. 18-1027, Col. Escandon, 11801 Mexico D.F. (Mexico)

2000-07-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Pr{sub 3}S{sub 2}Cl{sub 2}[AsS{sub 3}]: a praseodymium(III) sulfide chloride thioarsenate(III) with double chains of condensed [SPr{sub 4}]{sup 10+} tetrahedra

Energy Technology Data Exchange (ETDEWEB)

Kang, Dong-Hee [Institute for Inorganic Chemistry, University of Stuttgart (Germany); Department of Chemistry, University of Waterloo, ON (Canada); Korea Research Institute of Standards and Science (KRISS), Daejeon (Korea, Republic of); Ledderboge, Florian; Schleid, Thomas [Institute for Inorganic Chemistry, University of Stuttgart (Germany); Kleinke, Holger [Department of Chemistry, University of Waterloo, ON (Canada)

2015-02-15

Single crystals of Pr{sub 3}S{sub 2}Cl{sub 2}[AsS{sub 3}] were obtained by reacting praseodymium sesquisulfide and trichloride with arsenic and sulfur (molar ratio: 7:4:6:9) in torch-sealed evacuated silica ampoules at 850 C for a week and subsequent slow cooling to room temperature. The transparent, pale yellowish green platelets crystallize monoclinically (Pr{sub 3}S{sub 2}Cl{sub 2}[AsS{sub 3}]: a = 2190.96(9) pm, b = 685.49(3) pm, c = 701.87(3) pm, β = 98.752(3) ) in the non-centrosymmetric space group Cc with Z = 4. The crystal structure of Pr{sub 3}S{sub 2}Cl{sub 2}[AsS{sub 3}] comprises three crystallographically different Pr{sup 3+} cations. (Pr1){sup 3+} and (Pr2){sup 3+} exhibit bicapped trigonal prismatic environments of five sulfur and three chlorine atoms (C.N. = 8) each, whereas only seven sulfur atoms arranged as capped trigonal prism (C.N. = 7) are present in the case of (Pr3){sup 3+}. Two of the five crystallographically different sulfur atoms (S1 and S2) are coordinated by four Pr{sup 3+} cations, each in the shape of slightly distorted [SPr{sub 4}]{sup 10+} tetrahedra (d(S-Pr) = 278-294 pm). Hence, the crystal structure of Pr{sub 3}S{sub 2}Cl{sub 2}[AsS{sub 3}] contains double chains with the composition {sup 1}{sub ∞}{[S_2Pr_3]"5"+}, consisting of two parallel edge-fused chains of anti-SiS{sub 2}-analogous single strands {sup 1}{sub ∞}{[SPr_2]"4"+}, formed by trans-edge sharing [SPr{sub 4}]{sup 10+} tetrahedra. The arrangement of the {sup 1}{sub ∞}{[S_2Pr_3]"5"+} double chains running along [001] can be described as a hexagonal rod-packing. Chloride anions (Cl1 and Cl2) as well as the isolated, ψ{sup 1}-tetrahedral thioarsenate(III) groups [AsS{sub 3}]{sup 3-} (d(As-S) = 228-229 pm) consisting of three sulfur atoms (S3, S4, and S5) and an As{sup 3+} cation carrying a stereochemically active non-bonding electron pair take care of the three-dimensional interconnection and the charge balance of the {sup 1}{sub ∞}{[S_2Pr_3]"5"+} double

Continuity of care of emergency surgical admissions: impact on SpR training.

Science.gov (United States)

Ledwidge, S F C; Bryden, E; Halestrap, P; Galland, R B

2008-06-01

Continuity of patient care is an important component of surgical education. This study assesses continuity of care in the current working climate. Data were collected prospectively on consecutive emergency general surgical admissions during one month. Our SpR rota is a partial shift 24 hour on call with the SpR's own consultant. The SpR is free of commitments the next day following post-take work. The on call general surgery SpR was designated the 'assessor'. Data were analysed according to involvement of the 'assessor' at subsequent stages of the admission--consent, operation, review during admission and review on discharge. Data were also collected defining whether the 'assessor' and operator followed-up the patient. There were 200 admissions; 108 female and 92 male. Overall 23% admissions had the same 'assessor' for all stages of patient care. The 'assessor' dealt with an aspect of patient care in 11% of admissions who underwent an operation and 29% of admissions who were conservatively managed. SpR follow-up of admissions on whom they operated was 70% but only 41% of admissions who were conservatively managed were followed-up by the assessing SpR. Complete in-hospital continuity of care was poor, although SpR follow-up of patients on whom they had operated was better. Introduction of shift patterns has reduced continuity of patient care. This will have a negative impact on both surgical training and patient care.

«[…] en enkel og saklig innføring i norsk språkhistorie» Språkhistorie som norskdidaktisk emne – et faghistorisk utsyn

Directory of Open Access Journals (Sweden)

Stian Hårstad

2016-09-01

Full Text Available Artikkelen tar for seg språkhistorie som norskdidaktisk emne i et faghistorisk perspektiv. På grunnlag av en lang rekke styringsdokumenter (læreplaner og studieplaner og fagdidaktisk og disiplinfaglig litteratur undersøkes hvordan språkets historiske dimensjoner har blitt tematisert som kunnskapsområde. Et hovedfunn i denne gjennomgangen er at det historiske (eller historiserende lengdesnittet har vært til stede gjennom hele 1900-tallet og fortsatt finnes i gjeldende læreplan – om enn med temmelig varierende vektlegging og detaljering. Læreplanene har i stor grad overlatt til praktikerne (lærerne og lærerutdannerne å avgrense hva de språkhistoriske emnene skal romme. Litteraturanalysen viser videre at den didaktiske refleksjonen på dette feltet bare unntaksvis har kommet på trykk, og denne begrensete språkhistorie-didaktiske debatten må forstås på bakgrunn av tilstanden i forsknings- og studiefaget.Nøkkelord: språkhistorie-didaktikk, språkdidaktikk, språkhistorie, norskdidaktikkAbstractThis article deals with language history as a didactic topic in the subject Norwegian as L1. On the basis of a wide range of documents (curricula and syllabi and didactic and disciplinary literature, the article seeks to examine how the historical dimensions of language have been treated as an area of knowledge. A main finding is that the historic (or historicizing perspective has been present through the entire 20th century and still can be found in the current curriculum – though the emphasis and itemization have varied significantly. The curricula have largely entrusted the practicians (the teachers and teacher educators with the powers to delineate the language history topics. The research review further shows that the didactic reflections within this field only sporadically have appeared in print, and this limited didactic discussion of language history must be understood on the background of the condition of the discipline

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

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

1992-01-01

The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

Functional Conducting Polymers in the Application of SPR Biosensors

Directory of Open Access Journals (Sweden)

Rapiphun Janmanee

2012-01-01

Full Text Available In recent years, conducting polymers have emerged as one of the most promising transducers for both chemical, sensors and biosensors owing to their unique electrical, electrochemical and optical properties that can be used to convert chemical information or biointeractions into electrical or optical signals, which can easily be detected by modern techniques. Different approaches to the application of conducting polymers in chemo- or biosensing applications have been extensively studied. In order to enhance the application of conducting polymers into the area of biosensors, one approach is to introduce functional groups, including carboxylic acid, amine, sulfonate, or thiol groups, into the conducting polymer chain and to form a so-called “self-doped” or by doping with negatively charged polyelectrolytes. The functional conducting polymers have been successfully utilized to immobilize enzymes for construction of biosensors. Recently, the combination of SPR and electrochemical, known as electrochemical-surface plasmon resonance (EC-SPR, spectroscopy, has been used for in situ investigation of optical and electrical properties of conducting polymer films. Moreover, EC-SPR spectroscopy has been applied for monitoring the interaction between biomolecules and electropolymerized conjugated polymer films in biosensor and immunosensor applications. In this paper, recent development and applications on EC-SPR in biosensors will be reviewed.

SPR 2017. Abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2017-05-15

The conference proceedings SPR 2017 include abstracts on the following issues: gastrointestinal radiography - inflammatory bowel diseases, cardiovascular CTA, general muscoskeletal radiology, muscoskeletal congenital development diseases, general pediatric radiology - chest, muscoskeletal imaging - marrow and infectious disorders, state-of-the-art body MR imaging, practical pediatric sonography, quality and professionalism, CT imaging in congenital heart diseases, radiographic courses, body MT techniques, contrast enhanced ultrasound, machine learning, forensic imaging, the radiation dos conundrum - reconciling imaging, imagining and managing, the practice of radiology, interventional radiology, neuroradiology, PET/MR.

Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Wu Yuanyuan; Liu Tiancai; Sun Wei

2012-01-01

The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

Specialist committee's review reports for experimental fast reactor JOYO' MK-III performance tests

International Nuclear Information System (INIS)

Yamashita, Kiyonobu; Okubo, Toshiyuki; Kamide, Hideki

2004-02-01

Performance tests (startup-physics tests and power elevation tests) were planed for experimental fast reactor 'JOYO' MK-III where irradiation performances were upgraded by power increase from 100 to 140 MW. The reactor safety committee of O-arai Engineering Center has established a specialist committee for 'JOYO' MK-III Performance Tests at the first meeting of 2003 on 23th. April 2003, to accomplish the tests successfully. Subjects of the specialist committee were reviews of following items covering a wide range. 1) Contents of modification works. 2) Reflections of functional test results to the plant and facilities. 3) Reflections of safety rule modification to instruction and manual for operation. 4) Quality assurances and pre-calculation for performance test. 5) Inspection plan and its results. 6) Adequacy of performance test plan. 7) Confirmation of performance test results. Before test-starts, the specialist committee has confirmed by reviewing the items from 1) to 6) based on explanations and documents of the Division of Experimental Reactor, that the test plan and pre-inspections are adequate. After the tests, the specialist committee had confirmed by reviewing the item 7) in the same way, that the each test result satisfies the corresponding criterion. The specialist committee has concluded from these review's results before and after the tests that the 'JOYO' MK-III Performance Tests were carried out appropriately. Besides, the first criticality of the JOYO MK-III was achieved on 2nd. July 2003, and the continuous full power operation was carried on 20th. Nov. 2003. Finally, all performance tests were completed by the pass of the last governmental pre-serviced inspection (dose rate measurement during the shut down condition). (author)

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor

International Nuclear Information System (INIS)

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

2000-01-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Features of Bayou Choctaw SPR caverns and internal structure of the salt dome.

Energy Technology Data Exchange (ETDEWEB)

Munson, Darrell E.

2007-07-01

The intent of this study is to examine the internal structure of the Bayou Choctaw salt dome utilizing the information obtained from graphical representations of sonar survey data of the internal cavern surfaces. Many of the Bayou Choctaw caverns have been abandoned. Some existing caverns were purchased by the Strategic Petroleum Reserve (SPR) program and have rather convoluted histories and complex cavern geometries. In fact, these caverns are typically poorly documented and are not particularly constructive to this study. Only two Bayou Choctaw caverns, 101 and 102, which were constructed using well-controlled solutioning methods, are well documented. One of these was constructed by the SPR for their use while the other was constructed and traded for another existing cavern. Consequently, compared to the SPR caverns of the West Hackberry and Big Hill domes, it is more difficult to obtain a general impression of the stratigraphy of the dome. Indeed, caverns of Bayou Choctaw show features significantly different than those encountered in the other two SPR facilities. In the number of abandoned caverns, and some of those existing caverns purchased by the SPR, extremely irregular solutioning has occurred. The two SPR constructed caverns suggest that some sections of the caverns may have undergone very regular solutioning to form uniform cylindrical shapes. Although it is not usually productive to speculate, some suggestions that point to the behavior of the Bayou Choctaw dome are examined. Also the primary differences in the Bayou Choctaw dome and the other SPR domes are noted.

UV-SPR biosensor for biomolecular interaction studies

Science.gov (United States)

Geiss, F. A.; Fossati, S.; Khan, I.; Gisbert Quilis, N.; Knoll, W.; Dostalek, J.

2017-05-01

UV surface plasmon resonance (SPR) for direct in situ detection of protein binding events is reported. A crossed relief aluminum grating was employed for diffraction coupling to surface plasmons as an alternative to more commonly used attenuated total reflection method. Wavelength interrogation of SPR was carried out by using transmission measurements in order to probe odorant-binding protein 14 (OBP14) of the honey bee (Apis mellifera). The native oxide layer on the top of an aluminum grating sensor chip allows for covalent coupling of protein molecules by using regular silane-based linkers. The probing of bound OBP14 protein at UV with confined field of surface plasmons holds potential for further studies of interaction with recently developed artificial fluorescent odorants.

Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR; Diseno del mecanismo para la conduccion de las barras de control del reactor Triga Mark III del ININ.

Energy Technology Data Exchange (ETDEWEB)

Franco C, A

1997-12-01

This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author).

Screening and analysis of beyond design basis accident of 49-2 SPR

International Nuclear Information System (INIS)

Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

2015-01-01

The beyond design basis accident was analyzed to ensure safe operation of 49-2 Swimming Pool Reactor (SPR) after design life. Because it's difficult to use PSA method, the unconditional assumed severe accidents were adopted to obtain a conservative result. The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS), horizontal channel rupture, core uncovering after shutdown and emergency response capacity. The results show that the core is safe in SBO ATWS, and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors. The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed. (authors)

A Simple Small Size and Low Cost Sensor Based on Surface Plasmon Resonance for Selective Detection of Fe(III

Directory of Open Access Journals (Sweden)

Nunzio Cennamo

2014-03-01

Full Text Available A simple, small size, and low cost sensor based on a Deferoxamine Self Assembled Monolayer (DFO-SAM and Surface Plasmon Resonance (SPR transduction, in connection with a Plastic Optical Fiber (POF, has been developed for the selective detection of Fe(III. DFO-SAM sensors based on appropriate electrochemical techniques can be frequently found in the scientific literature. In this work, we present the first example of a DFO-SAM sensor based on SPR in an optical fiber. The SPR sensing platform was realized by removing the cladding of a plastic optical fiber along half the circumference, spin coating a buffer of Microposit S1813 photoresist on the exposed core, and finally sputtering a thin gold film. The hydroxamate siderophore deferoxamine (DFO, having high binding affinity for Fe(III, is then used in its immobilized form, as self-assembled monolayer on the gold layer surface of the POF sensor. The results showed that the DFO-SAM-POF-sensor was able to sense the formation of the Fe(III/DFO complex in the range of concentrations between 1 μm and 50 μm with a linearity range from 0 to 30 μm of Fe(III. The selectivity of the sensor was also proved by interference tests.

A simple small size and low cost sensor based on surface plasmon resonance for selective detection of Fe(III).

Science.gov (United States)

Cennamo, Nunzio; Alberti, Giancarla; Pesavento, Maria; D'Agostino, Girolamo; Quattrini, Federico; Biesuz, Raffaela; Zeni, Luigi

2014-03-07

A simple, small size, and low cost sensor based on a Deferoxamine Self Assembled Monolayer (DFO-SAM) and Surface Plasmon Resonance (SPR) transduction, in connection with a Plastic Optical Fiber (POF), has been developed for the selective detection of Fe(III). DFO-SAM sensors based on appropriate electrochemical techniques can be frequently found in the scientific literature. In this work, we present the first example of a DFO-SAM sensor based on SPR in an optical fiber. The SPR sensing platform was realized by removing the cladding of a plastic optical fiber along half the circumference, spin coating a buffer of Microposit S1813 photoresist on the exposed core, and finally sputtering a thin gold film. The hydroxamate siderophore deferoxamine (DFO), having high binding affinity for Fe(III), is then used in its immobilized form, as self-assembled monolayer on the gold layer surface of the POF sensor. The results showed that the DFO-SAM-POF-sensor was able to sense the formation of the Fe(III)/DFO complex in the range of concentrations between 1 μm and 50 μm with a linearity range from 0 to 30 μm of Fe(III). The selectivity of the sensor was also proved by interference tests.

Modernization of the facilities of the TRIGA Mark III reactor of ININ; Modernizacion de las instalaciones del reactor TRIGA Mark III del ININ

Energy Technology Data Exchange (ETDEWEB)

Mendez T, D.; Flores C, J., E-mail: dario.mendez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

International Nuclear Information System (INIS)

Wang Shisheng; Rineiski, Andrei; Maschek, Werner

2006-01-01

Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF 2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF 3 , etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as

Surface Plasmon Resonance (SPR Phenomenon of the Oxidizing and Reducing Polypyrrole

Directory of Open Access Journals (Sweden)

Nurlaila Rajabiah

2016-12-01

Full Text Available Surface Plasmon Resonance (SPR phenomenon of the oxidizing and reducing polypyrrole (PPy have been observed using a modified Au/PPy Kretschmann configuration. The observation was carried out through simulation Winspall 3.02 software and compared with some experimental data refractive index (n, absorbance index (, and thickness (τ reported in other literatures with spectroscopy ellipsometry. This simulation assumed that the SPR system use BK-7 halfcylinder prism (n= 1,515 and the laser beam was generated by HeNe (λ= 632,8 nm. The result showed that the optimum layer thickness of polypyrrole with neutral electrolyte solution in the reduction state is of about 20 nm. The polypyrrole with an acid solution in the oxidation state showed that the reflectivity curve is sharper and the width of the curve is smaller than the neutral electrolyte solution in the reduction state with the SPR angle of about 46,810 and the reflectivity value of about 0,217. Polypyrrole in a state of oxidation and reduction, related to the absorption, film thickness, and dielectric constant of materials, affected to the SPR angle shift and dip curve

Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor; Control difuso adaptable de la potencia neutronica del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Rojas R, E.

2014-07-01

The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n{sub o} up to a predetermined final level n{sub f}. During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

Reactor oscillator - I - III, Part III - Electronic device

International Nuclear Information System (INIS)

Lolic, B.; Jovanovic, S.

1961-12-01

This report describes functioning of the reactor oscillator electronic system. Two methods of oscillator operation were discussed. The first method is so called method of amplitude modulation of the reactor power, and the second newer method is phase method. Both methods are planned for the present reactor oscillator

Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR

International Nuclear Information System (INIS)

Franco C, A.

1997-01-01

This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author)

SPR based immunosensor for detection of Legionella pneumophila in water samples

Science.gov (United States)

Enrico, De Lorenzis; Manera, Maria G.; Montagna, Giovanni; Cimaglia, Fabio; Chiesa, Maurizio; Poltronieri, Palmiro; Santino, Angelo; Rella, Roberto

2013-05-01

Detection of legionellae by water sampling is an important factor in epidemiological investigations of Legionnaires' disease and its prevention. To avoid labor-intensive problems with conventional methods, an alternative, highly sensitive and simple method is proposed for detecting L. pneumophila in aqueous samples. A compact Surface Plasmon Resonance (SPR) instrumentation prototype, provided with proper microfluidics tools, is built. The developed immunosensor is capable of dynamically following the binding between antigens and the corresponding antibody molecules immobilized on the SPR sensor surface. A proper immobilization strategy is used in this work that makes use of an important efficient step aimed at the orientation of antibodies onto the sensor surface. The feasibility of the integration of SPR-based biosensing setups with microfluidic technologies, resulting in a low-cost and portable biosensor is demonstrated.

Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

International Nuclear Information System (INIS)

Chen Yiping; Liu Songlin

2010-01-01

Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

Characterization of the TRIGA Mark III reactor for k0-neutron activation analysis

International Nuclear Information System (INIS)

Diaz R, O.; Herrera P, E.; Lopez R, M.C.

1997-01-01

The non-ideality of the epithermal neutron flux distribution in a a reactor site parameter (α), the thermal-to-epithermal neutron ratio (f), the irradiation channel neutron temperature (T n ) and the k 0 -factors for more than 20 isotopes were determined in the 3 typical irradiation positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using different experimental methods with conventional and non-conventional monitors. This characterization is used in the k 0 -method of NAA, recently introduced at the Institute. (author). 21 refs., 3 figs., 5 tabs

Tritium system design for the mirror reactors FPD-I, FPD-II, and FPD-III

International Nuclear Information System (INIS)

Finn, P.A.

1985-01-01

The tritium system design for the Fusion Power Demonstration Reactor (FPD-I, II, and III) is described. The device operates at 25% availability. For FPD-II, an engineering mode using tritium neutral beams is part of the design

Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

Energy Technology Data Exchange (ETDEWEB)

Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

1972-07-01

This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

Dehumidification effects in the superheated region (SPR) of a direct expansion (DX) air cooling coil

International Nuclear Information System (INIS)

Xia Liang; Chan Mingyin; Deng Shiming; Xu Xiangguo

2009-01-01

A DX air cooling coil may normally be assumed to have two regions in its refrigerant side, according to refrigerant status, a two-phase region (TPR) and a superheated region (SPR). Dry air side surface of the SPR in a DX air cooling coil has been normally assumed in lumped-parameter mathematical models previously developed without however being validated. Therefore, an experimental study has been carried out to examine such an assumption under different operating conditions. The experimental results suggested that the air side surface of the SPR in a DX air cooling coil was either fully or partially wet under all experimental conditions and assuming dry air side of the SPR could lead to an underestimated total amount of water vapor condensed on the entire DX coil surface. Therefore, it is recommended that the assumption of dry air side in a SPR be no longer used in future lumped-parameter models to be developed for improved modeling accuracy.

Berkeley Nuclear Laboratories Reactor Physics Mk. III Experimental Programme. Description of facility and programme for 1971

Energy Technology Data Exchange (ETDEWEB)

Nunn, R M; Waterson, R H; Young, J D

1971-01-15

Reactor physics experiments have been carried out at Berkeley Nuclear Laboratories during the past few years in support of the Civil Advanced Gas-Cooled Reactors (Mk. II) the Generating Board is building. These experiments are part of an overall programme whose objective is to assess the accuracy of the calculational methods used in the design and operation of these reactors. This report provides a description of the facility for the Mk. III experimental programme and the planned programme for 1971.

Enhancement in sensitivity of graphene-based zinc oxide assisted bimetallic surface plasmon resonance (SPR) biosensor

Science.gov (United States)

Kumar, Rajeev; Kushwaha, Angad S.; Srivastava, Monika; Mishra, H.; Srivastava, S. K.

2018-03-01

In the present communication, a highly sensitive surface plasmon resonance (SPR) biosensor with Kretschmann configuration having alternate layers, prism/zinc oxide/silver/gold/graphene/biomolecules (ss-DNA) is presented. The optimization of the proposed configuration has been accomplished by keeping the constant thickness of zinc oxide (32 nm), silver (32 nm), graphene (0.34 nm) layer and biomolecules (100 nm) for different values of gold layer thickness (1, 3 and 5 nm). The sensitivity of the proposed SPR biosensor has been demonstrated for a number of design parameters such as gold layer thickness, number of graphene layer, refractive index of biomolecules and the thickness of biomolecules layer. SPR biosensor with optimized geometry has greater sensitivity (66 deg/RIU) than the conventional (52 deg/RIU) as well as other graphene-based (53.2 deg/RIU) SPR biosensor. The effect of zinc oxide layer thickness on the sensitivity of SPR biosensor has also been analysed. From the analysis, it is found that the sensitivity increases significantly by increasing the thickness of zinc oxide layer. It means zinc oxide intermediate layer plays an important role to improve the sensitivity of the biosensor. The sensitivity of SPR biosensor also increases by increasing the number of graphene layer (upto nine layer).

Optimization of SPR signals: Monitoring the physical structures and refractive indices of prisms

Science.gov (United States)

Maisarah Mukhtar, Wan; Halim, Razman Mohd; Hassan, Hazirah

2017-11-01

Surface plasmon resonance (SPR) can only be achieved if sufficient energy is provided at the boundary between metal and dielectric. An employment of prism as a light coupler by using Kretschmann configuration is one of the alternative for the production of adequate energy to be generated as surface plasmon polaritons (SPP). This work is carried out to investigate the effect of physical structure of the prism and its refractive index to the excitation of SPPs. A 50nm gold thin metal film with dielectric constant of ɛ=-12.45i+1.3 was deposited on the hypotenuse surface of the prisms. The physical structures of the prisms were varied such as triangular, conical, hemispherical and half cylindrical. These prisms were classified into two types of refractive indices (RI), namely n=1.51(type BK7) and n=1.77(type SF11). Based on SPR curve analyses, we discovered that strong SPR signals which consist of 82.98% photons were excited as SPPs can be obtained by using type-BK7 prism with physical structures of hemispherical or half cylindrical. From the view of selectivity ability as sensors, the usage of type-SF11 prisms (half cylindrical and hemispherical) able to enhance this impressive feature in which sharp SPR curves with small FWHM values were obtained. In conclusion, apart from properties of thin film materials, the physical structure of prisms and their RI values play crucial roles to obtain optimum SPR signal. High sensitivity SPR sensor can be established with the appointment of type-BK7 prisms (hemispherical or half cylindrical shape) as light couplers.

Optimization of SPR signals: Monitoring the physical structures and refractive indices of prisms

Directory of Open Access Journals (Sweden)

Mukhtar Wan Maisarah

2017-01-01

Full Text Available Surface plasmon resonance (SPR can only be achieved if sufficient energy is provided at the boundary between metal and dielectric. An employment of prism as a light coupler by using Kretschmann configuration is one of the alternative for the production of adequate energy to be generated as surface plasmon polaritons (SPP. This work is carried out to investigate the effect of physical structure of the prism and its refractive index to the excitation of SPPs. A 50nm gold thin metal film with dielectric constant of ɛ=-12.45i+1.3 was deposited on the hypotenuse surface of the prisms. The physical structures of the prisms were varied such as triangular, conical, hemispherical and half cylindrical. These prisms were classified into two types of refractive indices (RI, namely n=1.51(type BK7 and n=1.77(type SF11. Based on SPR curve analyses, we discovered that strong SPR signals which consist of 82.98% photons were excited as SPPs can be obtained by using type-BK7 prism with physical structures of hemispherical or half cylindrical. From the view of selectivity ability as sensors, the usage of type-SF11 prisms (half cylindrical and hemispherical able to enhance this impressive feature in which sharp SPR curves with small FWHM values were obtained. In conclusion, apart from properties of thin film materials, the physical structure of prisms and their RI values play crucial roles to obtain optimum SPR signal. High sensitivity SPR sensor can be established with the appointment of type-BK7 prisms (hemispherical or half cylindrical shape as light couplers.

Språkpolitikk og (usynleggjering i det semiotiske landskapet på Universitetet i Tromsø

Directory of Open Access Journals (Sweden)

Åse Mette Johansen

2013-01-01

Full Text Available I denne artikkelen analyserer vi det semiotiske landskapet som famnar om og er ein del av staden Universitetet i Tromsø (UiT, inkludert Universitetssykehuset Nord-Norge (UNN. Analysen bygger på eit todelt datamateriale som om­fattar fotografi frå universitets­området i tillegg til relevante dokument om språk­lovgjeving og andre offisielle vedtak om språkbruk. Dei aktuelle språka er norsk (bokmål og nynorsk, nord­samisk, engelsk og kvensk/finsk. I tillegg blir ymse slag visuelle symbol under­kasta analyse. Også kunstnarleg utsmykking blir kort omtala. Vi viser korleis synleggjering av ulike språk på UiT er eit resultat av språkpolitikk på ulike nivå, men også er påverka av meir implisitte faktorar som lokale og nasjonale språkhierarki og, ikkje minst, konstruksjonen av ein regional institusjons­identitet som blir uttrykt gjennom det semiotiske landskapet meir generelt. Som ramme rundt denne analysen gjer vi dessutan ein ekskurs til det som i 2011 utløyste ein intens debatt i Tromsø kommune, nemleg spørsmålet om kommunen burde eller ikkje burde bli innlemma i forvaltningsområdet for samisk språk. Den måten denne problemstillinga er blitt takla på av politiske organ i kommunen, og ikkje minst debatten om den i medieoffentlegheita, står i skarp kontrast til den fleirspråklege freden som omgjev universitetet og universitets­sjukehuset.

Experimental results for characterization of a tapered plastic optical fiber sensor based on SPR

Science.gov (United States)

Cennamo, N.; Galatus, R.; Zeni, L.

2015-05-01

The experimental results obtained with two different Plastic Optical Fiber (POF) geometries, tapered and not-tapered, for a sensor based on Surface Plasmon Resonance (SPR) are presented. SPR is used for determining the refractive index variations at the interface between a gold layer and a dielectric medium (aqueous medium). In this work SPR sensors in POF configurations, useful for bio-sensing applications, have been realized for the optimization of the sensitivity and experimentally tested. The results show as the sensitivity increases with the tapered POF configuration, when the refractive index of aqueous medium increases.

SPR sensors for monitoring the degradation processes of Eu(dbm)3(phen) and Alq3 thin films under atmospheric and UVA exposure

Science.gov (United States)

Del Rosso, T.; Zaman, Q.; Cremona, M.; Pandoli, O.; Barreto, A. R. J.

2018-06-01

The degradation processes of tris(8-hydroxyquinoline) (Alq3) and tris(dibenzoylmethane) mono(1,10-phenanthroline)europium(III) (Eu(dbm)3(phen)) thin films are investigated by the use of AFM, photoluminescence and SPR spectroscopy. The plasmonic sensors are operated both in air and nitrogen environments, where they are irradiated with controlled doses of UVA radiation. AFM results don't reveal the formation of heterogeneous phases and crystallization under air exposure. The organic thin films change their refractive index under both types of exposure and act as a protective layer against oxidation for the SiO2/MPTS/metal interface of the plasmonic sensors. SPR measurements reveal a strict correlation between the refractive index increase and quenching of the photoluminescence of the organic thin films. The results are promising for the development of compact plasmonic UVA dosimeters in the surface plasmon coupled emission configuration (SPCE) with lanthanide β-diketonate complex materials (patent pending).

Effekten af sprøjtemiddelafdrift på buske og træer i læhegn

DEFF Research Database (Denmark)

Kjær, C.; Strandberg, M. T.; Erlandsen, M.

Forsøg med sprøjtning af tjørn med ukrudtsmidlet metsul-furon-methyl har vist effekter både i samme år som sprøjtning og året efter. Sprøjtemidlet, som anvendes til bekæmpelse af bredbladet ukrudt i kornmarker, kan formentlig ved afdrift til markomgivelser som hegn og skovkanter medføre en længer...

Safety in the ARIES-III D-3He tokamak reactor design

International Nuclear Information System (INIS)

Herring, J.S.; Dolan, T.J.

1992-01-01

This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

Energy Technology Data Exchange (ETDEWEB)

Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

2012-07-01

The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

Two-Channel SPR Sensor Combined Application of Polymer- and Vitreous-Clad Optic Fibers

Directory of Open Access Journals (Sweden)

Yong Wei

2017-12-01

Full Text Available By combining a polymer-clad optic fiber and a vitreous-clad optic fiber, we proposed and fabricated a novel optic fiber surface plasmon resonance (SPR sensor to conduct two-channel sensing at the same detection area. The traditional optic fiber SPR sensor has many disadvantages; for example, removing the cladding requires corrosion, operating it is dangerous, adjusting the dynamic response range is hard, and producing different resonance wavelengths in the sensing area to realize a multi-channel measurement is difficult. Therefore, in this paper, we skillfully used bare fiber grinding technology and reverse symmetry welding technology to remove the cladding in a multi-mode fiber and expose the evanescent field. On the basis of investigating the effect of the grinding angle on the dynamic range change of the SPR resonance valley wavelength and sensitivity, we combined polymer-clad fiber and vitreous-clad fiber by a smart design structure to realize at a single point a two-channel measurement fiber SPR sensor. In this paper, we obtained a beautiful spectral curve from a multi-mode fiber two-channel SPR sensor. In the detection range of the refractive rate between 1.333 RIU and 1.385 RIU, the resonance valley wavelength of channel Ⅰ shifted from 622 nm to 724 nm with a mean average sensitivity of 1961 nm/RIU and the resonance valley wavelength of channel Ⅱ shifted from 741 nm to 976 nm with a mean average sensitivity of 4519 nm/RIU.

Two-Channel SPR Sensor Combined Application of Polymer- and Vitreous-Clad Optic Fibers.

Science.gov (United States)

Wei, Yong; Su, Yudong; Liu, Chunlan; Nie, Xiangfei; Liu, Zhihai; Zhang, Yu; Zhang, Yonghui

2017-12-09

By combining a polymer-clad optic fiber and a vitreous-clad optic fiber, we proposed and fabricated a novel optic fiber surface plasmon resonance (SPR) sensor to conduct two-channel sensing at the same detection area. The traditional optic fiber SPR sensor has many disadvantages; for example, removing the cladding requires corrosion, operating it is dangerous, adjusting the dynamic response range is hard, and producing different resonance wavelengths in the sensing area to realize a multi-channel measurement is difficult. Therefore, in this paper, we skillfully used bare fiber grinding technology and reverse symmetry welding technology to remove the cladding in a multi-mode fiber and expose the evanescent field. On the basis of investigating the effect of the grinding angle on the dynamic range change of the SPR resonance valley wavelength and sensitivity, we combined polymer-clad fiber and vitreous-clad fiber by a smart design structure to realize at a single point a two-channel measurement fiber SPR sensor. In this paper, we obtained a beautiful spectral curve from a multi-mode fiber two-channel SPR sensor. In the detection range of the refractive rate between 1.333 RIU and 1.385 RIU, the resonance valley wavelength of channel Ⅰ shifted from 622 nm to 724 nm with a mean average sensitivity of 1961 nm/RIU and the resonance valley wavelength of channel Ⅱ shifted from 741 nm to 976 nm with a mean average sensitivity of 4519 nm/RIU.

Licensing of the TRIGA Mark III reactor at the Mexican Nuclear Centre

International Nuclear Information System (INIS)

Ramirez, R.M.; Arrendondo, R.R.

1990-01-01

The TRIGA Mark III reactor at the Mexican Nuclear Centre went critical in 1968 and remained so until 1979 when the National Commission for Nuclear Safety and Safeguards (CNSNS), the Mexican regulatory authority, was set up. The reactor was therefore operating without a formal operating license, and the CNSNS accordingly requested the ININ to license the reactor under the existing conditions and to ensure that any modification of the original design complied with Standards ANSI/ANS-15 and with the code of practice set out in IAEA Safety Series No. 35. The most relevant points in granting the operating licence were: (a) the preparation of the Safety Report; (b) the formulation and application of the Quality Assurance Programme; (c) the reconditioning of the following reactor systems: the cooling systems; the ventilation and exhaust system; the monitoring system and control panel; (d) the training of the reactor operating staff at junior and senior levels; and (e) the formulation of procedures and instructions. Once the provisional operating license was obtained for the reactor it was considered necessary to modify the reactor core, which has been composed of 20% enriched standards fuel, to a mixed core based on a mixture of standard fuel and FLIP-type fuel with 70% 235 U enrichment. The CNSNS therefore requested that the mixed core be licensed and a technical report was accordingly annexed to the Safety Report, its contents including the following subjects: (a) neutron analysis of the proposed configuration; (b) reactor shutdown margins; (c) accident analysis; and (d) technical specifications. The licensing process was completed this year and we are now hoping to obtain the final operating license

Combinations of SPR and MS for Characterizations of Native and Recombinant Proteins in Cell Lysates

DEFF Research Database (Denmark)

Borch, Jonas; Roepstorff, Peter

2006-01-01

Surface plasmon resonance and mass spectrometry (SPR-MS) has been combined for quality check of recombinant 6xHis-tagged 14-3-3 proteins expressed in Escherichia coli. Lysates were injected over an SPR sensorchip with immobilized Ni2+ for SPR analysis of the specific Ni2+ binding response...... and stability. To validate the identity, intactness and homogeneity of the captured proteins were eluted for mass spectrometric analysis of intact molecular weight and peptide mass mapping. Additionally, the captured recombinant proteins were investigated for specific binding to known phosphorylated ligands...... of 14-3-3 proteins in order to test their activity. Specific binding of recombinant and native 14-3-3 proteins in complex mixtures to immobilized phosphopeptides and subsequent elution was also tested by SPR-MS. Ammonium sulfate precipitate fractions from lysates of E. coli expressing 14-3-3 protein...

Enhanced antibody recognition with a magneto-optic surface plasmon resonance (MO-SPR) sensor.

Science.gov (United States)

Manera, Maria Grazia; Ferreiro-Vila, Elías; Garcia-Martin, José Miguel; Garcia-Martin, Antonio; Rella, Roberto

2014-08-15

A comparison between sensing performance of traditional SPR (Surface Plasmon Resonance) and magneto-optic SPR (MOSPR) transducing techniques is presented in this work. MOSPR comes from an evolution of traditional SPR platform aiming at modulating Surface Plasmon wave by the application of an external magnetic field in transverse configuration. Previous work demonstrated that, when the Plasmon resonance is excited in these structures, the external magnetic field induces a modification of the coupling of the incident light with the Surface Plasmon Polaritons (SPP). Besides, these structures can lead to an enhancement in the magneto-optical (MO) activity when the SPP is excited. This phenomenon is exploited in this work to demonstrate the possibility to use the enhanced MO signal as proper transducer signal for investigating biomolecular interactions in liquid phase. To this purpose, the transducer surface was functionalized by thiol chemistry and used for recording the binding between Bovine Serum Albumin molecules immobilized onto the surface and its complementary target. Higher sensing performance in terms of sensitivity and lower limit of detection of the MOSPR biosensor with respect to traditional SPR sensors is demonstrated. Copyright © 2014 Elsevier B.V. All rights reserved.

Správa nemovitosti versus facility management

OpenAIRE

Rázga, Štěpán

2008-01-01

Problematiku facility managementu a správy nemovitostí práce uceleně shrnuje a porovnává teoretické předpoklady a metodické postupy plynoucí z výuky facility managementu na VŠE v Praze s výkonem daných činností v praxi.

Pleiotropic Regulation of Virulence Genes in Streptococcus mutans by the Conserved Small Protein SprV.

Science.gov (United States)

Shankar, Manoharan; Hossain, Mohammad S; Biswas, Indranil

2017-04-15

Streptococcus mutans , an oral pathogen associated with dental caries, colonizes tooth surfaces as polymicrobial biofilms known as dental plaque. S. mutans expresses several virulence factors that allow the organism to tolerate environmental fluctuations and compete with other microorganisms. We recently identified a small hypothetical protein (90 amino acids) essential for the normal growth of the bacterium. Inactivation of the gene, SMU.2137, encoding this protein caused a significant growth defect and loss of various virulence-associated functions. An S. mutans strain lacking this gene was more sensitive to acid, temperature, osmotic, oxidative, and DNA damage-inducing stresses. In addition, we observed an altered protein profile and defects in biofilm formation, bacteriocin production, and natural competence development, possibly due to the fitness defect associated with SMU.2137 deletion. Transcriptome sequencing revealed that nearly 20% of the S. mutans genes were differentially expressed upon SMU.2137 deletion, thereby suggesting a pleiotropic effect. Therefore, we have renamed this hitherto uncharacterized gene as sprV ( s treptococcal p leiotropic r egulator of v irulence). The transcript levels of several relevant genes in the sprV mutant corroborated the phenotypes observed upon sprV deletion. Owing to its highly conserved nature, inactivation of the sprV ortholog in Streptococcus gordonii also resulted in poor growth and defective UV tolerance and competence development as in the case of S. mutans Our experiments suggest that SprV is functionally distinct from its homologs identified by structure and sequence homology. Nonetheless, our current work is aimed at understanding the importance of SprV in the S. mutans biology. IMPORTANCE Streptococcus mutans employs several virulence factors and stress resistance mechanisms to colonize tooth surfaces and cause dental caries. Bacterial pathogenesis is generally controlled by regulators of fitness that are

Real-time ligation chain reaction for DNA quantification and identification on the FO-SPR.

Science.gov (United States)

Knez, Karel; Spasic, Dragana; Delport, Filip; Lammertyn, Jeroen

2015-05-15

Different assays have been developed in the past years to meet point-of-care diagnostic tests requirements for fast and sensitive quantification and identification of targets. In this paper, we developed the ligation chain reaction (LCR) assay on the Fiber Optic Surface Plasmon Resonance (FO-SPR) platform, which enabled simultaneous quantification and cycle-to-cycle identification of DNA during amplification. The newly developed assay incorporated FO-SPR DNA melting assay, previously developed by our group. This required establishment of several assay parameters, including buffer ionic strength and thermal ramping speed as these parameters both influence the ligation enzyme performance and the hybridization yield of the gold nanoparticles (Au NPs) on the FO-SPR sensor. Quantification and identification of DNA targets was achieved over a wide concentration range with a calibration curve spanning 7 orders of magnitude and LOD of 13.75 fM. Moreover, the FO-SPR LCR assay could discriminate single nucleotide polymorphism (SNPs) without any post reaction analysis, featuring thus all the essential requirements of POC tests. Copyright © 2014 Elsevier B.V. All rights reserved.

Reactor oscillator - I - III, Part I

International Nuclear Information System (INIS)

Lolic, B.

1961-12-01

Project 'Reactor oscillator' covers the following activities: designing reactor oscillators for reactors RA and RB with detailed engineering drawings; constructing and mounting of the oscillator; designing and constructing the appropriate electronic equipment for the oscillator; measurements at the RA and RB reactors needed for completing the oscillator construction

Neutron spectra in two beam ports of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Aguilar, F.; Paredes, L.; Rivera M, T.

2013-10-01

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Neutron detection of the Triga Mark III reactor, using nuclear track methodology

Energy Technology Data Exchange (ETDEWEB)

Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I. [Instituto de Física, Universidad Nacional Autónoma de México Circuito de la Investigación Científica, Ciudad Universitaria. México, DF (Mexico); Raya-Arredondo, R.; Cruz-Galindo, S. [Instituto Nacional de Investigaciones Nucleares (Mexico); Sajo-Bohus, L. [Universidad Simón Bolivar, Laboratorio de Física Nuclear, Caracas (Venezuela, Bolivarian Republic of)

2015-07-23

Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

Neutron spectra in two beam ports of the TRIGA Mark III reactor

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

2013-10-15

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

International Nuclear Information System (INIS)

In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers

BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

1981-06-01

This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

1981-06-01

This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

A molecular imprinted SPR biosensor for sensitive determination of citrinin in red yeast rice.

Science.gov (United States)

Atar, Necip; Eren, Tanju; Yola, Mehmet Lütfi

2015-10-01

A novel and sensitive molecular imprinted surface plasmon resonance (SPR) biosensor was developed for selective determination of citrinin (CIT) in red yeast rice. Firstly, the gold surface of SPR chip was modified with allyl mercaptane. Then, CIT-imprinted poly(2-hydroxyethyl methacrylate-methacryloylamidoglutamic acid) (p(HEMA-MAGA)) film was generated on the gold surface modified with allyl mercaptane. The unmodified and imprinted surfaces were characterized by Fourier transform infrared (FTIR) spectroscopy, atomic force microscopy (AFM) and contact angle measurements. The linearity range and the detection limit were obtained as 0.005-1.0 ng/mL and 0.0017 ng/mL, respectively. The SPR biosensor was applied to determination of CIT in red yeast rice sample. Copyright © 2015 Elsevier Ltd. All rights reserved.

Enhanced magneto-optical SPR platform for amine sensing based on Zn porphyrin dimers

OpenAIRE

Manera, M. G.; Ferreiro-Vila, E.; Garcia-Martin, J. M.; Cebollada, A.; Garcia-Martin, A.; Giancane, G.; Valli, L.; Rella, R.

2013-01-01

Ethane-bridged Zn porphyrins dimers (ZnPP) have been deposited by Langmuir-Schäfer (LS) deposition technique onto proper transducer layers for surface plasmon resonance (SPR) and magneto-optical surface plasmon resonance (MO-SPR) characterization techniques performed in controlled atmosphere. This last tool has emerged as a novel and very performing sensing technique using as transducer layers a combination of noble and magnetic layers deposited onto glass substrates. A magnetic actuation all...

Reactor oscillator - I - III, Part III - Electronic device; Reaktorski oscilator - I-III, III Deo - Elektronski uredjaj

Energy Technology Data Exchange (ETDEWEB)

Lolic, B; Jovanovic, S [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report describes functioning of the reactor oscillator electronic system. Two methods of oscillator operation were discussed. The first method is so called method of amplitude modulation of the reactor power, and the second newer method is phase method. Both methods are planned for the present reactor oscillator.

Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Rojas R, E.

2014-01-01

The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n o up to a predetermined final level n f . During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

Assessments of the probabilities of aircraft impact with the Sandia Pulsed Reactor and Building 836, Sandia Laboratories, Albuquerque

International Nuclear Information System (INIS)

Biringer, B.E.

1976-11-01

This report documents a study of the annual probabilities of aircraft impact with the Sandia Pulsed Reactor (SPR) and Bldg. 836 at Sandia Laboratories, Albuquerque. The probability of aircraft impact into each structure was estimated using total yearly operations, effective structure area, structure location relative to air activity, and accident rate per kilometer. The estimated probability for an aircraft impact with SPR is 1.1 x 10 -4 per year; the estimated probability for impact with Bldg. 836 is 1.0 x 10 -3 per year

Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Gold nanoparticle-enhanced multiplexed imaging surface plasmon resonance (iSPR) detection of Fusarium mycotoxins in wheat

Science.gov (United States)

A rapid, sensitive and multiplexed imaging surface plasmon resonance (iSPR) biosensor assay was developed and validated for three Fusarium toxins, deoxynivalenol (DON), zearalenone (ZEA) and T-2 toxin. The iSPR assay was based on a competitive inhibition format with secondary antibodies (Ab2) conjug...

Theoretical analysis of nuclear reactors (Phase III), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, III Deo, Zatrovanje reaktora

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-01-15

Report on calculation of poisoning in experimental and power reactor includes four parts. Part one describes the influence of poisoning on the physical parameters of a reactor. part two includes transformation of differential equations for iodine and xenon. It was needed for easier solution of of differential equation using the analog computer. This calculation was done for RA reactor operating at 5 MW power. The RA reactor was used an example of calculation by the proposed method. Part four shows the application of the method for calculating the Calder Hall power reactor.

Effect of sulfasalazine on human neuroblastoma: analysis of sepiapterin reductase (SPR) as a new therapeutic target

International Nuclear Information System (INIS)

Yco, Lisette P.; Geerts, Dirk; Mocz, Gabor; Koster, Jan; Bachmann, André S.

2015-01-01

Neuroblastoma (NB) is an aggressive childhood malignancy in children up to 5 years of age. High-stage tumors frequently relapse even after aggressive multimodal treatment, and then show therapy resistance, typically resulting in patient death. New molecular-targeted compounds that effectively suppress tumor growth and prevent relapse with more efficacy are urgently needed. We and others previously showed that polyamines (PA) like spermidine and spermine are essential for NB tumorigenesis and that DFMO, an inhibitor of the key PA synthesis gene product ODC, is effective both in vitro and in vivo, securing its evaluation in NB clinical trials. To find additional compounds interfering with PA biosynthesis, we tested sulfasalazine (SSZ), an FDA-approved salicylate-based anti-inflammatory and immune-modulatory drug, recently identified to inhibit sepiapterin reductase (SPR). We earlier presented evidence for a physical interaction between ODC and SPR and we showed that RNAi-mediated knockdown of SPR expression significantly reduced native ODC enzyme activity and impeded NB cell proliferation. Human NB mRNA expression datasets in the public domain were analyzed using the R2 platform. Cell viability, isobologram, and combination index analyses as a result of SSZ treatment with our without DFMO were carried out in NB cell cultures. Molecular protein-ligand docking was achieved using the GRAMM algorithm. Statistical analyses were performed with the Kruskal-Wallis test, 2log Pearson test, and Student’s t test. In this study, we show the clinical relevance of SPR in human NB tumors. We found that high SPR expression is significantly correlated to unfavorable NB characteristics like high age at diagnosis, MYCN amplification, and high INSS stage. SSZ inhibits the growth of NB cells in vitro, presumably due to the inhibition of SPR as predicted by computational docking of SSZ into SPR. Importantly, the combination of SSZ with DFMO produces synergistic antiproliferative effects

Effect of sulfasalazine on human neuroblastoma: analysis of sepiapterin reductase (SPR) as a new therapeutic target.

Science.gov (United States)

Yco, Lisette P; Geerts, Dirk; Mocz, Gabor; Koster, Jan; Bachmann, André S

2015-06-21

Neuroblastoma (NB) is an aggressive childhood malignancy in children up to 5 years of age. High-stage tumors frequently relapse even after aggressive multimodal treatment, and then show therapy resistance, typically resulting in patient death. New molecular-targeted compounds that effectively suppress tumor growth and prevent relapse with more efficacy are urgently needed. We and others previously showed that polyamines (PA) like spermidine and spermine are essential for NB tumorigenesis and that DFMO, an inhibitor of the key PA synthesis gene product ODC, is effective both in vitro and in vivo, securing its evaluation in NB clinical trials. To find additional compounds interfering with PA biosynthesis, we tested sulfasalazine (SSZ), an FDA-approved salicylate-based anti-inflammatory and immune-modulatory drug, recently identified to inhibit sepiapterin reductase (SPR). We earlier presented evidence for a physical interaction between ODC and SPR and we showed that RNAi-mediated knockdown of SPR expression significantly reduced native ODC enzyme activity and impeded NB cell proliferation. Human NB mRNA expression datasets in the public domain were analyzed using the R2 platform. Cell viability, isobologram, and combination index analyses as a result of SSZ treatment with our without DFMO were carried out in NB cell cultures. Molecular protein-ligand docking was achieved using the GRAMM algorithm. Statistical analyses were performed with the Kruskal-Wallis test, 2log Pearson test, and Student's t test. In this study, we show the clinical relevance of SPR in human NB tumors. We found that high SPR expression is significantly correlated to unfavorable NB characteristics like high age at diagnosis, MYCN amplification, and high INSS stage. SSZ inhibits the growth of NB cells in vitro, presumably due to the inhibition of SPR as predicted by computational docking of SSZ into SPR. Importantly, the combination of SSZ with DFMO produces synergistic antiproliferative effects

Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor; Desarrollo de la interface de usuario para la visualizacion de los sistemas auxiliares del reactor nuclear Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Merced D, J. E.

2016-07-01

The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

Surface functionalization of SPR chip for specific molecular interaction analysis under flow condition

Directory of Open Access Journals (Sweden)

Tao Ma

2017-03-01

Full Text Available Surface functionalization of sensor chip for probe immobilization is crucial for the biosensing applications of surface plasmon resonance (SPR sensors. In this paper, we report a method circulating the dopamine aqueous solution to coat polydopamine film on sensing surface for surface functionalization of SPR chip. The polydopamine film with available thickness can be easily prepared by controlling the circulation time and the biorecognition elements can be immobilized on the polydopamine film for specific molecular interaction analysis. These operations are all performed under flow condition in the fluidic system, and have the advantages of easy implementation, less time consuming, and low cost, because the reagents and devices used in the operations are routinely applied in most laboratories. In this study, the specific absorption between the protein A probe immobilized on the sensing surface and human immunoglobulin G in the buffer is monitored based on this surface functionalization strategy to demonstrated its feasibility for SPR biosensing applications.

Thermal spectra of the TRIGA Mark III reactor; El espectro termico del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Macias B, L.R.; Palacios G, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1998-07-01

The diffraction phenomenon is gave in observance of the well known Bragg law in crystalline materials and this can be performance by mean of X-rays, electrons and neutrons among others, which allows to do inside the field of each one of these techniques the obtaining of measurements focussed at each one of them. For the present work, it will be mentioned only the referring to X-ray and neutron techniques. The X-ray diffraction due to its properties just it does measurements which are known in general as superficial measurements of the sample material but for the properties of the neutrons, this diffraction it explores in volumetric form the sample material. Since the neutron diffraction process depends lots of its intensity, then it is important to know the neutron source spectra that in this case is supplied by the TRIGA Mark III reactor. Within of diffraction techniques a great number of them can be found, however some of the traditional will be mentioned such as the identification of crystalline samples, phases identification and the textures measurement. At present this last technique is founded on the dot of a minimum error and the technique of phases identification performs but not compete with that which is obtained by mean of X-rays due to this last one has a major resolution. (Author)

Modernization of the facilities of the TRIGA Mark III reactor of ININ

International Nuclear Information System (INIS)

Mendez T, D.; Flores C, J.

2016-09-01

The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia M, H.; Emeterio H, M.; Canizal S, C. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, C.P. 11801 Mexico D.F. (Mexico)

2000-07-01

This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor; Elaboracion de la configuracion y programacion del sistema de interbloqueos del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Mejia C, M. A.

2016-07-01

The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

International Working Group on Past Reactors Thirteenth Annual Meeting. Summary Report. Part III

International Nuclear Information System (INIS)

1981-04-01

The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

Fatigue strength of a single lap joint SPR-bonded

International Nuclear Information System (INIS)

Di Franco, G.; Fratini, L.; Pasta, A.

2011-01-01

In the last years, hybrid joints, meaning with this the joints which consist in combining a traditional mechanical joint to a layer of adhesive, are gradually attracting the attention of various sectors of the construction of vehicles and transportation industries, for their better performance compared to just mechanical joints (self-piercing riveting SPR, riveting, and so on) or just to bonded joints.The paper investigates the fatigue behavior of a single lap joint self-piercing riveted (SPR) and bonded throughout fatigue tests. The considered geometric configuration allowed the use of two rivets placed longitudinally; an epoxy resin was used as adhesive. In the first part of the work static characterization of the joints was carried out through tensile tests. Then fatigue tests were made with the application of different levels of load. The fatigue curves were also obtained at the varying the distance between the two rivets in order to better assess the joint strength for a given length of overlap.

Analysis of SPR salt cavern remedial leach program 2013.

Energy Technology Data Exchange (ETDEWEB)

Weber, Paula D.; Gutierrez, Karen A.; Lord, David L.; Rudeen, David Keith

2013-09-01

The storage caverns of the US Strategic Petroleum Reserve (SPR) exhibit creep behavior resulting in reduction of storage capacity over time. Maintenance of oil storage capacity requires periodic controlled leaching named remedial leach. The 30 MMB sale in summer 2011 provided space available to facilitate leaching operations. The objective of this report is to present the results and analyses of remedial leach activity at the SPR following the 2011 sale until mid-January 2013. This report focuses on caverns BH101, BH104, WH105 and WH106. Three of the four hanging strings were damaged resulting in deviations from normal leach patterns; however, the deviations did not affect the immediate geomechanical stability of the caverns. Significant leaching occurred in the toes of the caverns likely decreasing the number of available drawdowns until P/D ratio criteria are met. SANSMIC shows good agreement with sonar data and reasonably predicted the location and size of the enhanced leaching region resulting from string breakage.

Special awards lighten up SPR banquet

Science.gov (United States)

Zwickl, Ron; Baker, Dan; Reiff, Pat

On December 8, 1982, Marcia Neugebauer, then president of the Solar and Planetary Relationships section, held the “First Occasional Awards Ceremony” in conjunction with the annual dinner banquet at the Fall AGU meeting. These awards were an attempt to add a little humor to our usually somber gatherings. This year we reincarnated Neugebauer's successful concept and presented a number of new and novel awards to our fellow scientists at the Fall AGU SPR dinner. Summarized for your enjoyment are the categories and the official winners, as announced at the December 6 banquet.

Reactor oscillator - I - III, Part I; Reaktorski oscilator - I-III, I Deo

Energy Technology Data Exchange (ETDEWEB)

Lolic, B [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Project 'Reactor oscillator' covers the following activities: designing reactor oscillators for reactors RA and RB with detailed engineering drawings; constructing and mounting of the oscillator; designing and constructing the appropriate electronic equipment for the oscillator; measurements at the RA and RB reactors needed for completing the oscillator construction.

Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC; Coordinacion de los sistemas motrices de las barras de control del reactor Triga Mark III, mediante el uso de un PLC

Energy Technology Data Exchange (ETDEWEB)

Castro R, P. R.

2016-07-01

The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor; Algoritmo neuro-difuso para la identificacion de la potencia neutronica del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Rojas R, E.; Benitez R, J. S. [Instituto Tecnologico de Toluca, Division de Estudios de Posgrado e Investigacion, Av. Tecnologico s/n, Ex-Rancho La Virgen, 50140 Metepec, Estado de Mexico (Mexico); Segovia de los Rios, J. A.; Rivero G, T. [ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jorge.benitez@inin.gob.mx

2009-10-15

In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

The SPR detection of Salmonella enteritidis in food using aptamers as recongnition elements

Science.gov (United States)

Di, W. T.; Du, X. W.; Pan, M. F.; Wang, J. P.

2017-09-01

In this experiment, a fast, accurate, non-destructive, unmarked and simple-operation detection method for Salmonella enteritidis in food was established by the BI-3000 plasma resonance biosensor (SPR). This article establishes a method of using nucleic acid aptamer as immune recognition element in SPR which can be employed to detect Salmonella enteritidis in food for the first time. The experimental conditions were screened and the experimental scheme was validated and applied. The best flow rate was 5μL/min, the best concentration of the aptamers was 180mM, and the best regenerating solution was the 20mM NaOH. This method had almost no cross-reactivity. Besides, we established a standard curve of Salmonella enteritidis and SPR signal, with the detection limit of 2 cfu/mL. Finally, we tested the samples of chicken, pork, shrimp and fish purchased from supermarkets. The method has the advantages of short time, low detection limit and easy operation, which can be used for a large number of food samples.

Floating Chip Mounting System Driven by Repulsive Force of Permanent Magnets for Multiple On-Site SPR Immunoassay Measurements

Science.gov (United States)

Horiuchi, Tsutomu; Tobita, Tatsuya; Miura, Toru; Iwasaki, Yuzuru; Seyama, Michiko; Inoue, Suzuyo; Takahashi, Jun-ichi; Haga, Tsuneyuki; Tamechika, Emi

2012-01-01

We have developed a measurement chip installation/removal mechanism for a surface plasmon resonance (SPR) immunoassay analysis instrument designed for frequent testing, which requires a rapid and easy technique for changing chips. The key components of the mechanism are refractive index matching gel coated on the rear of the SPR chip and a float that presses the chip down. The refractive index matching gel made it possible to optically couple the chip and the prism of the SPR instrument easily via elastic deformation with no air bubbles. The float has an autonomous attitude control function that keeps the chip parallel in relation to the SPR instrument by employing the repulsive force of permanent magnets between the float and a float guide located in the SPR instrument. This function is realized by balancing the upward elastic force of the gel and the downward force of the float, which experiences a leveling force from the float guide. This system makes it possible to start an SPR measurement immediately after chip installation and to remove the chip immediately after the measurement with a simple and easy method that does not require any fine adjustment. Our sensor chip, which we installed using this mounting system, successfully performed an immunoassay measurement on a model antigen (spiked human-IgG) in a model real sample (non-homogenized milk) that included many kinds of interfering foreign substances without any sample pre-treatment. The ease of the chip installation/removal operation and simple measurement procedure are suitable for frequent on-site agricultural, environmental and medical testing. PMID:23202030

Development and validation of a model TRIGA Mark III reactor with code MCNP5

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ; Caracterizacion del flujo neutronico en el Hohlraum de la columna termica del reactor TRIGA Mark III del ININ

Energy Technology Data Exchange (ETDEWEB)

Delfin L, A.; Palacios, J.C.; Alonso, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

2006-07-01

Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor

International Nuclear Information System (INIS)

Merced D, J. E.

2016-01-01

The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

Detection of mycotoxins using imaging surface plasmon resonance (iSPR)

Science.gov (United States)

Significant progress has been made in the development of biosensors that can be used to detect mycotoxins. One technology that has been extensively tested is surface plasmon resonance (SPR). In 2003 a multi-toxin method was reported that detected aflatoxin B1 (AFB1), zearalenone (ZEA), fumonisin B1 ...

Estimation of optical constants of a bio-thin layer (onion epidermis), using SPR spectroscopy

International Nuclear Information System (INIS)

Rehman, Saif-ur-; Hayashi, Shinji; Sekkat, Zouheir; Mumtaz, Huma; Shaukat, S F

2014-01-01

We estimate the optical constants of a biological thin layer (Allium cepa) by surface plasmon resonance (SPR) spectroscopy. For this study, the fresh inner thin epidermis of an onion bulb was used and stacked directly on gold (Au) and silver (Ag) film surfaces in order to identify the shift in SPR mode of each metal film at an operating wavelength of 632.8 nm. The thickness and dielectric constants of the biological thin layer were determined by matching the experimental SPR curves to theoretical ones. The thickness and roughness of bare Au and Ag thin films were also measured by atomic force microscopy (AFM); the results of which are in good agreement with those obtained through experiment. Due to the high surface roughness of the natural onion epidermis layer, AFM could not measure the exact thickness of an onion epidermis. It is estimated that the value of the real part of the dielectric constant of an onion epidermis is between the dielectric constants of water and air. (paper)

Enhancement of chitosan-graphene oxide SPR sensor with a multi-metallic layers of Au–Ag–Au nanostructure for lead(II) ion detection

Energy Technology Data Exchange (ETDEWEB)

Kamaruddin, Nur Hasiba [Department of Electric, Electronic and Systems Engineering, Faculty of Engineering and Built Environment, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia); Bakar, Ahmad Ashrif A., E-mail: ashrif@ukm.edu.my [Department of Electric, Electronic and Systems Engineering, Faculty of Engineering and Built Environment, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia); Yaacob, Mohd Hanif; Mahdi, Mohd Adzir [Wireless and Photonic Network Research Centre, Faculty of Engineering, Universiti Putra Malaysia, 43400 Serdang, Selangor (Malaysia); Zan, Mohd Saiful Dzulkefly [Department of Electric, Electronic and Systems Engineering, Faculty of Engineering and Built Environment, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia); Shaari, Sahbudin [Institute of Microengineering and Nanoelectronics, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia)

2016-01-15

Highlights: • Tri-metallic Au–Ag–Au CS-GO SPR sensor was fabricated for the first time. • The tri-metallic nanostructure provided an enhanced evanescent field. • Successful functionalization of the CS-GO sensing layer. • Superior performance for lead(II) ion detection. - Abstract: We demonstrate the enhancement of surface plasmon resonance (SPR) technique by implementing a multi-metallic layers of Au–Ag–Au nanostructure in the chitosan-graphene oxide (CS-GO) SPR sensor for lead(II) ion detection. The performance of the sensor is analyzed via SPR measurements, from which the sensitivity, signal-to-noise ratio and repeatability are determined. The nanostructure layers are characterized using field emission scanning electron microscopy (FESEM), atomic force microscopy (AFM), X-ray diffraction (XRD), Raman spectroscopy and X-ray photoelectron spectroscopy (XPS). We showed that the proposed structure has increased the shift in the SPR angle up to 3.5° within the range of 0.1–1 ppm due to the enhanced evanescent field at the sensing layer-analyte interface. This sensor also exhibits great repeatability which benefits from the stable multi-metallic nanostructure. The SNR value of 0.92 for 5 ppm lead(II) ion solution and reasonable linearity range up to that concentration shows that the tri-metallic CS-GO SPR sensor gives a good response towards the lead(II) ion solution. The CS-GO SPR sensor is also sensitive to at least a 10{sup −5} change in the refractive index. The results prove that our proposed tri-metallic CS-GO SPR sensor demonstrates a strong performance and reliability for lead(II) ion detection in accordance with the standardized lead safety level for wastewater.

Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

2009-10-01

In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC

International Nuclear Information System (INIS)

Castro R, P. R.

2016-01-01

The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

Prepare for an SpR interview.

Science.gov (United States)

Garner, J P; Heppell, P S J

2003-03-01

By the time you attend an interview for a military SpR number you should have no real problems but it pays to be prepared. Begin preparations early, reading widely and talk to as many people as possible. Your consultants will have a useful viewpoint on the proceedings and may be able to help you refine your answers to the common questions. Arrive at your interview in a smart and timely fashion and answer questions with confidence and common sense. Avoid confrontation and bluff and be courteous at all times, whatever you may be feeling inside and thank the interview panel as you leave.

Binding behavior of CRP and anti-CRP antibody analyzed with SPR and AFM measurement

International Nuclear Information System (INIS)

Lee, Soo-Keun; Kim, Hyun-Chul; Cho, Sang-Joon; Jeong, Sang Won; Jeon, Won Bae

2008-01-01

Atomic force microscope (AFM) was exploited to take picture of the molecular topology of C-reactive protein (CRP) in phosphate-buffered saline (PBS) solution. An explicit molecular image of CRP demonstrated a pentagonal structure composed of five subunits. Dimensions of the doughnut-shaped CRP molecule measured by AFM were about 25 nm in outside diameter and 10 nm in central pore diameter, and the height of CRP molecule was about 4 nm which was comparable to the value determined by X-ray crystallography. Bis(N-succinimido)-11,11'-dithiobis (undecyl succinate) (DSNHS) was synthesized for use as a linker for immobilizing anti-CRP antibody (anti-CRP) onto the gold surface of a surface plasmon resonance (SPR) sensor chip. DSNHS formed self-assembled monolayer (SAM) on the gold surface. By use of an AFM tip, a pattern of ditch was engraved within the SAM of DSNHS, and anti-CRP was immobilized on the engraved SAM through replacement of N-hydroxysuccinimide group on the outside surface of DSNHS by the amine group of anti-CRP. Formation of CRP/anti-CRP complex on the gold surface of SPR sensor chip was clearly demonstrated by measuring SPR angle shift. A consecutive series of SAM, SAM/anti-CRP, and SAM/anti-CRP/CRP complexes was generated on a SPR sensor chip, and the changes in depth of the ditch were monitored by taking AFM images of the complexes. Comparative analysis of the depth differences indicates that binding of CRP to anti-CRP occurs in a planar mode

Development of surface plasmon resonance (SPR) biosensors for use in the diagnostics of malignant and infectious diseases

Science.gov (United States)

Firdous, S.; Anwar, S.; Rafya, R.

2018-06-01

Surface plasmon resonance (SPR) has become an important optical biosensing technology due to its real-time, label-free, and noninvasive nature. These techniques allow for rapid and ultra-sensitive detection of biological analytes, with applications in medical diagnostics, environmental monitoring, and agriculture. SPR is widely used in the detection of biomolecular interactions, and improvements are required for both sensitivity and in vivo uses for practical applications. In this study, we developed an SPR biosensor to provide a highly sensitive and specific approach to early-stage detection of viral and malignant diseases, such as cancer tumors, for which biomarker detection is very important. A cancer cell line (HeLa cells) with biomarker Rodamine 6G was experimentally analyzed in vitro with our constructed SPR biosensor. It was observed that the biosensor can offer a potentially powerful solution for tumor screening with dominant angular shift. The angular shift for both regents is dominant with a time curve at a wavelength of 632.8 nm of a He–Ne laser. We have successfully captured and detected a biomarker in vitro for cancer diagnostics using the developed instrument.

Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

Experimental measurement of the refrigerant temperature of the TRIGA Mark III reactor of the ININ

International Nuclear Information System (INIS)

Gallardo S, L.F.; Alonso V, G.

1991-08-01

With the object of knowing the axial temperature profile of the refrigerant in the core of the TRIGA Mark III reactor of the ININ, the temperatures of this, at the enter, in the center and the exit of the core were measured, in the positions: west 2, north 2 and south 1. This was made by means of the thermo pars introduction mounted in aluminum guides, connected to a measurer of digital temperature, whose resolution is of ± 0.1 C. The measurements showed a bigger heating of the refrigerant in the superior half of the core, that which suggests that the axial profile of temperature of the reactor is not symmetrical with respect to the center or that those temperature measurements in the center are not correct. (Author)

Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-III

International Nuclear Information System (INIS)

Suzuki, Tohru; Tobita, Yoshiharu; Kondo, Satoru; Saito, Yasushi; Mishima, Kaichiro

2003-01-01

SIMMER-III, a safety analysis code for liquid-metal fast reactors (LMFRs), includes a momentum exchange model based on conventional correlations for ordinary gas-liquid flows, such as an air-water system. From the viewpoint of safety evaluation of core disruptive accidents (CDAs) in LMFRs, we need to confirm that the code can predict the two-phase flow behaviors with high liquid-to-gas density ratios formed during a CDA. In the present study, the momentum exchange model of SIMMER-III was assessed and improved using experimental data of two-phase flows containing liquid metal, on which fundamental information, such as bubble shapes, void fractions and velocity fields, has been lacking. It was found that the original SIMMER-III can suitably represent high liquid-to-gas density ratio flows including ellipsoidal bubbles as seen in lower gas fluxes. In addition, the employment of Kataoka-Ishii's correlation has improved the accuracy of SIMMER-III for gas-liquid metal flows with cap-shape bubbles as identified in higher gas fluxes. Moreover, a new procedure, in which an appropriate drag coefficient can be automatically selected according to bubble shape, was developed. Through this work, the reliability and the precision of SIMMER-III have been much raised with regard to bubbly flows for various liquid-to-gas density ratios

The Development of a Portable SPR Bioanalyzer for Sensitive Detection of Escherichia coli O157:H7

Directory of Open Access Journals (Sweden)

Shun Wang

2016-11-01

Full Text Available The purpose of this study was to develop a portable surface plasmon resonance (SPR bioanalyzer for the sensitive detection of Escherichia coli O157:H7 in comparison with an enzyme-linked immunosorbent assay (ELISA. The experimental setup mainly consisted of an integrated biosensor and a homemade microfluidic cell with a three-way solenoid valve. In order to detect Escherichia coli O157:H7 using the SPR immunoassay, 3-mercaptopropionic acid (3-MPA was chemisorbed onto a gold surface via covalent bond for the immobilization of biological species. 1-ethyl-3-(3-dimethylaminopropyl carbodiimide hydrochloride (EDC and N-hydroxysuccinimide (NHS were used as crosslinker reagents to enable the reaction between 3-MPA and Escherichia coli O157:H7 antibodies by covalent –CO–NH– amide bonding. The experimental results were obtained from the Escherichia coli O157:H7 positive samples prepared by 10-, 20-, 40-, 80-, and 160-fold dilution respectively, which show that a good linear relationship with the correlation coefficient R of 0.982 existed between the response units from the portable SPR bioanalyzer and the concentration of Escherichia coli O157:H7 positive samples. Moreover, the theoretical detection limit of 1.87 × 103 cfu/mL was calculated from the positive control samples. Compared with the Escherichia coli O157:H7 ELISA kit, the sensitivity of this portable SPR bioanalyzer is four orders of magnitude higher than the ELISA kit. The results demonstrate that the portable SPR bioanalyzer could provide an alternative method for the quantitative and sensitive determination of Escherichia coli O157:H7 in field.

Status of research reactors in China. Their utilization and safety upgrading

International Nuclear Information System (INIS)

Xu Hanming; Jin Huajin

2000-01-01

The main research reactors in China basically consist of several old reactors including HWRR, HFETR, SPR, MJTR and MNSR. Except the last one, all the other reactors operate at a high power density and represent themselves as main tools in China for engineering testing, radioactive isotope production, and neutron scattering research. The research and production activities by these reactors are briefed. Main equipment and research topics for neutron scattering are described. The production of radioisotope is summarized. Safety upgrading activities in recent years taken by these old reactors are described, which make the safety feature of each reactor significantly improved and on the whole more close to (even not completely consistent) with the targets set by the modern safety regulation. Since a new multi-purpose research reactor CARR is expected available around the year of 2005, a schedule about the construction of new reactor, reforming or decommissioning of old reactors and smoothly transition of research and production activities from old to new reactor during the coming years has been under careful planning. A suggestion of potential international cooperation items has been preliminarily given. (author)

LUSH-based SPR sensor for the detection of alcohols and pheromone

Science.gov (United States)

Lau, Hui-Chong; Lee, Yeon-Kyung; Kwon, Jae-Young; Sohn, Young-Soo; Lim, Jeong Ok

2013-05-01

Protein is a widely used sensing substrate in the biosensing technology. In the study conducted here, we used odorant binding protein, LUSH from Drosophila as a biosensing substrate in a miniaturized surface plasmon resonance (SPR) sensor. LUSH contains the specific alcohols binding sites, which mediates the detection of alcohols and pheromone. We first modified the surface of the gold sensor chip using the self assembled monolayer in the chloroform solution. The saturated concentration was determined prior to the detection of alcohols and pheromone at various concentrations. The results showed that the LUSH was saturated at 1000 μg/ml on the gold sensor chip. The detection response of LUSH was significant at higher concentration of alcohols. LUSH detected ethanol at concentration >=50% propanol was detected at >=25% whereas pheromone was detected at >=1.25 μg/μl. The results provide some fundamental information on the potential use of LUSH-based SPR as a simple and easy protein-based sensor in the near future.

The behavior of crude oil spot and futures prices around OPEC and SPR announcements: An event study perspective

International Nuclear Information System (INIS)

Demirer, Riza; Kutan, Ali M.

2010-01-01

This paper examines the informational efficiency of crude oil spot and futures markets with respect to OPEC conference and U.S. Strategic Petroleum Reserve (SPR) announcements. We employ the event study methodology to examine the abnormal returns in crude oil spot and futures markets around OPEC conference and SPR announcement dates between 1983 and 2008. Our findings regarding OPEC announcements indicate an asymmetry in that only OPEC production cut announcements yield a statistically significant impact with the impact diminishing for longer maturities. We also find that the persistence of returns following OPEC production cut announcements creates substantial excess returns to investors who take long positions on the day following the end of OPEC conferences. In the case of SPR announcements, we find that the government's use of this program initiates a short-run market reaction following the announcement date. Furthermore, our tests of cumulative abnormal returns suggest that the market reacts efficiently to SPR announcements providing support for the use of the strategic reserves as a tool to stabilize the oil market. Our findings have significant policy implications for investors and are useful in designing effective energy policy strategies.

Comparison of the CPU and memory performance of StatPatternRecognitions (SPR) and Toolkit for MultiVariate Analysis (TMVA)

International Nuclear Information System (INIS)

Palombo, G.

2012-01-01

High Energy Physics data sets are often characterized by a huge number of events. Therefore, it is extremely important to use statistical packages able to efficiently analyze these unprecedented amounts of data. We compare the performance of the statistical packages StatPatternRecognition (SPR) and Toolkit for MultiVariate Analysis (TMVA). We focus on how CPU time and memory usage of the learning process scale versus data set size. As classifiers, we consider Random Forests, Boosted Decision Trees and Neural Networks only, each with specific settings. For our tests, we employ a data set widely used in the machine learning community, “Threenorm” data set, as well as data tailored for testing various edge cases. For each data set, we constantly increase its size and check CPU time and memory needed to build the classifiers implemented in SPR and TMVA. We show that SPR is often significantly faster and consumes significantly less memory. For example, the SPR implementation of Random Forest is by an order of magnitude faster and consumes an order of magnitude less memory than TMVA on Threenorm data.

The nuclear reactor systems

International Nuclear Information System (INIS)

Bacher, P.

2008-01-01

This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

Vurdering af omfang og konsekvenser af sprøjtning og gødskning af § 3-beskyttede naturarealer

DEFF Research Database (Denmark)

Strandberg, Morten Tune; Bak, Jesper Leth; Bladt, Jesper Stentoft

2012-01-01

Det er overvejende de ferske enge, der gødskes og sprøjtes, hvorimod dette kun undtagelsesvist er konstateret på andre naturtyper som overdrev og heder m.m. De ferske enge er samtidig den mest udbredte § 3-naturtype. Behandling i form af gødskning og sprøjtning foregår anslået på op til 20 % af a...

Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

2014-01-01

Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

Considerations Related to LTO for Gen II/III NPP

International Nuclear Information System (INIS)

Cojan, Mihail

2012-01-01

Today there are some 435 nuclear power reactors operating in 30 countries with a combined capacity of over 372 GWe. In 2011 these provided 2518 billion kWh, about 14% of the world's electricity [1]. The next evolutionary design of Generation III reactors to be deployed over many decades will represent a large part of the worldwide fleet throughout the 21st century. Generation III reactors are the future NPPs with improved safety and reliability, with passive safety systems and with a very low probability for core melt. The objective of this paper is to present the R and D activities that support LTO for Generation II / III Nuclear Reactors. (author)

The application and design of distributed control system in reactor shutdown system of Qinshan phase III

International Nuclear Information System (INIS)

Su Guoquan; Liu Wangtian; Yu Yijun; Xiong Weihua

2006-03-01

The design, commissioning and running of the reactor trip parameter monitoring system used in Qinshan Phase III are introduced. The applying technology of Distributed Control System realized trip parameter monitoring and realized the function of trip parameters quick data acquisitioning, transferring, saving, alarm, query. The applying of trip parameters monitoring system improved the abilities of plant status monitoring and event analyzing, and increased the security and economy of nuclear power plant. (authors)

Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor

International Nuclear Information System (INIS)

Garcia M, H.; Emeterio H, M.; Canizal S, C.

1999-01-01

This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

Summary report of the experimental fast reactor JOYO MK-III performance test

International Nuclear Information System (INIS)

Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro

2004-03-01

An upgrading project (MK-III project) was started to improve the irradiation capability of the experimental fast reactor JOYO. In this project, core replacement and increase of the reactor thermal power by the factor 1.4 were necessary for increasing the maximum fast neutron flux by the factor 1.3 and doubling the capacity for irradiation rigs. The modification of the cooling system that included the replacement of the main intermediate heat exchangers and the dump heat exchangers was completed in September 2000. After a series of system function tests, the performance test, of which objective is to fully characterize the upgraded core and heat transfer system, was started in June 2003. Twenty eight tests were selected and carried out as performance test, in order to confirm that the whole plant satisfy the design criteria and have sufficient characteristics (data necessary for safe and steady operation, core management, reactor control and monitoring) as an irradiation bed. After attaining the initial criticality of the core on 2nd July 2003, core characteristics (the excess reactivity, the isotherm temperature reactivity coefficient, the power reactivity coefficient and so on), plant characteristics (the plant heat balance, the adjustment of the temperature control system, the plant behavior at transient), shielding characteristics (dose rate distribution). As the result, it was confirmed that all the criteria regulated was satisfied and the core and plant have sufficient margins for full power operation, which was increased by the factor 1.4. Especially, nuclear analysis accuracy was verified by comparing the calculation with measured core characteristics of the initial core which consists of fifty five fresh fuel subassemblies. The operational data which is supposed to be useful for developing in-core anomaly detection system were also obtained. The operation manual and training simulator and design of next reactor development were revised based on the results

Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Mejia C, M. A.

2016-01-01

The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

Epithermal neutron flux characterization of the TRIGA Mark III reactor, Salazar, Mexico, for use in Internal Neutron Activation Analysis

International Nuclear Information System (INIS)

Diaz Rizo, O.; Herrera Peraza, E.

1996-01-01

The non ideality of the epithermal neutron flux distribution at a reactor site parameter (made, using Chloramine-T method. Radiochemical purity and stability of the labelled product were determined by radiochromatography. The labelled Melagenine-II showed two radioactive fractions thermal-to-epithermal neutron ratio (f) were determined in the 3 typical irradiations positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using the Cd-ratio for multi monitor and bare bi-isotopic monitor methods respectively. This characterization is of use in the K o - method of neutron activation analysis, recently introduced at the Institute

Characterization of the irradiation facilities SINCA and SIRCA of the TRIGA Mark III reactor using the code MCNPX

International Nuclear Information System (INIS)

Delfin L, A.; Garcia M, T.; Lucatero, M. A.; Cruz G, H. S.; Gonzalez, J. A.; Vargas E, S.

2011-11-01

The commitment of changing fuels of high enrichment for fuels of low enrichment in the TRIGA Mark III reactor of the Nuclear Center of Mexico generates the necessity to know the distribution of the spectrum of the neutrons flux in the irradiation facilities like they are: the Pneumatic System of Capsules Irradiation and the Rotational System of Capsules Irradiation. Is very important for the experiments design as well as for the reactor safety to know the profiles of the neutrons flux and the spectrum that these maintain with the mixed core with which operates, to effect of conserving the same characteristics when the reactor core will be operated with fuel of low enrichment totally. Also, knowing the profiles of the neutrons flux, the reactor operators can optimize the irradiation conditions of the processed samples and likewise the users can select the irradiation positions more adaptable to their necessities. This work present the characterization of the neutron flux in the irradiation facilities SINCA and SIFCA, calculated with the code MCNPX. (Author)

Yellin's review of the Nuclear Regulatory Commission's Reactor Safety Study: comment

International Nuclear Information System (INIS)

Wilson, R.

1976-01-01

Joel Yellin (Bell J. Economics, Vol. 7, No. 1 (Spr. 1976)) reviewed the Reactor Safety Study (WASH-1400, or Rasmussen Report) published by the Nuclear Regulatory Commission and makes several criticisms. Wilson finds that some of these criticisms are much overstated, while others, although valid, are mainly criticisms of presentation. To be useful for public policy, reactor risks--and indeed all nuclear risks--must be compared with other risks society faces. Wilson feels that the Rasmussen Report is weak in these comparisons, primarily because there exist few estimates of risk in other places, but that the comparisons he does make indeed confirm that the risk is smaller than Yellin suggests. 20 references

Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

Activation calculation of steel of the control rods of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A.

2014-10-01

In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

SPR based hybrid electro-optic biosensor for β-lactam antibiotics determination in water

Science.gov (United States)

Galatus, Ramona; Feier, Bogdan; Cristea, Cecilia; Cennamo, Nunzio; Zeni, Luigi

2017-09-01

The present work aims to provide a hybrid platform capable of complementary and sensitive detection of β-lactam antibiotics, ampicillin in particular. The use of an aptamer specific to ampicillin assures good selectivity and sensitivity for the detection of ampicillin from different matrice. This new approach is dedicated for a portable, remote sensing platform based on low-cost, small size and low-power consumption solution. The simple experimental hybrid platform integrates the results from the D-shape surface plasmon resonance plastic optical fiber (SPR-POF) and from the electrochemical (bio)sensor, for the analysis of ampicillin, delivering sensitive and reliable results. The SPR-POF already used in many previous applications is embedded in a new experimental setup with fluorescent fibers emitters, for broadband wavelength analysis, low-power consumption and low-heating capabilities of the sensing platform.

Updated neutron spectrum characterization of SNL baseline reactor environments

International Nuclear Information System (INIS)

Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

1994-04-01

This document provides SAND-II and MANIPULATE output listings from calculations used to derive the new spectrum-averaged integral parameters that were reported in volume 1. When used in conjunction with volume 1, this document provides an audit trail for the neutron radiation field characterization and supports current quality assurance initiatives. This document provides detailed information on the neutron spectrum characteristics of the primary Sandia National Laboratories' (SNL) reactor environments. The information in this volume is not intended for the casual user of the SNL reactor facilities. This detailed characterization of the neutron and gamma environments at the Sandia Pulsed Reactor (SPR) and the Annular Core Research Reactor (ACRR) is provided to aid the users who wish to convert the information given in the Radiation Metrology Laboratory (RML) dosimetry reports into other (non-silicon) measures of neutron damage. The spectra provided in these appendices can be used as a source term for Monte Carlo radiation transport calculations to study the impact of experimenter's test package on the neutron environment

Evaluation of Gd{sub 2}O{sub 2}S:Pr granular phosphor properties for X-ray mammography imaging

Energy Technology Data Exchange (ETDEWEB)

David, S.; Michail, C. [Department of Biomedical Engineering, Technological Educational Institute (TEI) of Athens, Ag. Spyridonos Street, 122 10 Egaleo (Greece); Seferis, I. [Department of Biomedical Engineering, Technological Educational Institute (TEI) of Athens, Ag. Spyridonos Street, 122 10 Egaleo (Greece); Faculty of Chemistry, Wroclaw University, 14F Joliot-Curie Street, 50-383 Wroclaw (Poland); Valais, I.; Fountos, G.; Liaparinos, P.; Kandarakis, I. [Department of Biomedical Engineering, Technological Educational Institute (TEI) of Athens, Ag. Spyridonos Street, 122 10 Egaleo (Greece); Kalyvas, N., E-mail: nkalyvas@teiath.gr [Department of Biomedical Engineering, Technological Educational Institute (TEI) of Athens, Ag. Spyridonos Street, 122 10 Egaleo (Greece)

2016-01-15

Phosphor materials are widely used in X-ray medical imaging detector applications, coupled with suitable photoreceptors. Upon the most demanding imaging modality is X-ray mammography, since the best defense against breast cancer is its early detection. A material suitable as a mammographic detector should efficiently absorb X-ray photons and transform them to optical photons, so as to minimize breast dose. The aim of the present study was to investigate the X-ray absorption efficiency and the absolute efficiency (AE), defined as the output optical photon power divided by the incident exposure, of Gd{sub 2}O{sub 2}S:Pr powder scintillator. For the purposes of this study, three scintillating screens with coating thicknesses, 34.1, 46.0 and 81.5 mg/cm{sup 2} respectively, were prepared in our laboratory from Gd{sub 2}O{sub 2}S:Pr powder (Phosphor Technology, Ltd.) by sedimentation on silica substrates. The quantum detection efficiency (QDE), the energy absorption efficiency (EAE), the spectral matching factor and the absolute efficiency (AE) were evaluated for X-ray mammographic conditions. Furthermore theoretical models were utilized to investigate the optical photon transmission properties through the phosphor mass. Gd{sub 2}O{sub 2}S:Pr presented high X-ray absorption properties and good spectral compatibility with several photoreceptors. It may be utilized for X-ray mammographic imaging if it is put in conjunction with a sensitive photoreceptor, so as to enhance Gd{sub 2}O{sub 2}S:Pr light emission properties. - Highlights: • Gd{sub 2}O{sub 2}S:Pr phosphor evaluated for mammography detectors. • The X-ray absorption efficiency was found high. • Spectral matching compatibility found for several photoreceptors. • X-ray absolute efficiency measured smaller than other phosphors. • Optical diffusion length and the light transmission per layer was theoretically calculated.

SIMMER-III analytic thermophysical property model

International Nuclear Information System (INIS)

Morita, K; Tobita, Y.; Kondo, Sa.; Fischer, E.A.

1999-05-01

An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms are designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless steel, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-core materials. (author)

RB research nuclear reactor, Annual report for 1983, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.

1983-01-01

The annual report for 1981 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff; financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; utilization of the reactor as a radiation source. It contains the preliminary safety report for operating the reactor with the internal neutron converter and the plan for criticality experiment with the converter

RB research nuclear reactor, Annual report for 1984, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.; Ilic, I.

1984-01-01

The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source

RB research nuclear reactor - Annual report for 1986, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

1987-01-01

This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986 [sr

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-11-01

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO 3 ) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10 18 n/cm 2 , which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

Configuration-controlled Au nanocluster arrays on inverse micelle nano-patterns: versatile platforms for SERS and SPR sensors

Science.gov (United States)

Jang, Yoon Hee; Chung, Kyungwha; Quan, Li Na; Špačková, Barbora; Šípová, Hana; Moon, Seyoung; Cho, Won Joon; Shin, Hae-Young; Jang, Yu Jin; Lee, Ji-Eun; Kochuveedu, Saji Thomas; Yoon, Min Ji; Kim, Jihyeon; Yoon, Seokhyun; Kim, Jin Kon; Kim, Donghyun; Homola, Jiří; Kim, Dong Ha

2013-11-01

Nanopatterned 2-dimensional Au nanocluster arrays with controlled configuration are fabricated onto reconstructed nanoporous poly(styrene-block-vinylpyridine) inverse micelle monolayer films. Near-field coupling of localized surface plasmons is studied and compared for disordered and ordered core-centered Au NC arrays. Differences in evolution of the absorption band and field enhancement upon Au nanoparticle adsorption are shown. The experimental results are found to be in good agreement with theoretical studies based on the finite-difference time-domain method and rigorous coupled-wave analysis. The realized Au nanopatterns are exploited as substrates for surface-enhanced Raman scattering and integrated into Kretschmann-type SPR sensors, based on which unprecedented SPR-coupling-type sensors are demonstrated.Nanopatterned 2-dimensional Au nanocluster arrays with controlled configuration are fabricated onto reconstructed nanoporous poly(styrene-block-vinylpyridine) inverse micelle monolayer films. Near-field coupling of localized surface plasmons is studied and compared for disordered and ordered core-centered Au NC arrays. Differences in evolution of the absorption band and field enhancement upon Au nanoparticle adsorption are shown. The experimental results are found to be in good agreement with theoretical studies based on the finite-difference time-domain method and rigorous coupled-wave analysis. The realized Au nanopatterns are exploited as substrates for surface-enhanced Raman scattering and integrated into Kretschmann-type SPR sensors, based on which unprecedented SPR-coupling-type sensors are demonstrated. Electronic supplementary information (ESI) available: TEM image and UV-vis absorption spectrum of citrate-capped Au NPs, AFM images of Au NC arrays on the PS-b-P4VP (41k-24k) template, ImageJ-analyzed results of PS-b-P4VP (41k-24k)-templated Au NC arrays, calculated %-surface coverage values, SEM images of Au NC arrays on the PS-b-P2VP (172k-42k

Thermal spectra of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Macias B, L.R.; Palacios G, J.

1998-01-01

The diffraction phenomenon is gave in observance of the well known Bragg law in crystalline materials and this can be performance by mean of X-rays, electrons and neutrons among others, which allows to do inside the field of each one of these techniques the obtaining of measurements focussed at each one of them. For the present work, it will be mentioned only the referring to X-ray and neutron techniques. The X-ray diffraction due to its properties just it does measurements which are known in general as superficial measurements of the sample material but for the properties of the neutrons, this diffraction it explores in volumetric form the sample material. Since the neutron diffraction process depends lots of its intensity, then it is important to know the neutron source spectra that in this case is supplied by the TRIGA Mark III reactor. Within of diffraction techniques a great number of them can be found, however some of the traditional will be mentioned such as the identification of crystalline samples, phases identification and the textures measurement. At present this last technique is founded on the dot of a minimum error and the technique of phases identification performs but not compete with that which is obtained by mean of X-rays due to this last one has a major resolution. (Author)

Engineering design of ARIES-III

International Nuclear Information System (INIS)

Sze, D.K.; Wong, C.; Cheng, E.

1993-07-01

An efficient organic cooled low activation ferritic steel first wall and shield has been designed for the D- 3 He power reactor ARIES-III. The design allows removal of the large surface heat load without exceeding temperature and stress design limits. The structure is expected to last for the whole reactor life. The major concerns regarding using the organic coolant in fusion reactors have been greatly alleviated

Organic coolant for ARIES-III

International Nuclear Information System (INIS)

Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

1991-04-01

ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

Use of the foil activation method with arbitrary trial functions to determine neutron energy spectra

International Nuclear Information System (INIS)

Kelly, J.G.; Vehar, D.W.

1987-01-01

Neutron Spectra have been measured by the foil activation method in thirteen different environments in and around the Sandia Pulsed Reactor (SPR-III), the White Sands Missile Range FBR, and the Annular Core Research Reactor (ACRR). The unfolded spectra were obtained by using the SANDII code in a manner which was not dependent on the initial trial. This altered technique is, therefore, better suited for the determination of spectra in environments that are difficult to predict by calculation, and it tends to reveal features that may be biased out by the use of standard trial functions

Comparative SPR study on the effect of nanomaterials on the biological activity of adsorbed proteins

International Nuclear Information System (INIS)

Mei, Q.; Chen, Y.; Hong, J.; Chen, H.; Ding, X.; Yin, Y.; Koh, K.; Lee, J.

2012-01-01

Bioactivity of proteins is evaluated to test the adverse effects of nanoparticles interjected into biological systems. Surface plasmon resonance (SPR) spectroscopy detects binding affinity that is normally related to biological activity. Utilizing SPR spectroscopy, a concise testing matrix is established by investigating the adsorption level of bovine serum albumin (BSA) and anti-BSA on the surface covered with 11-mercaptoundecanoic acid (MUA); magnetic nanoparticles (MNPs) and single-walled carbon nanotubes (SWCNTs), respectively. The immunoactivity of BSA on MNPs and SWCNT decreased by 18 % and 5 %, respectively, compared to that on the gold film modified with MUA. This indicates that MNPs cause a considerable loss of biological activity of adsorbed protein. This effect can be utilized for practical applications on detailed biophysical research and nanotoxicity studies. (author)

Stability Modification of SPR Silver Nano-Chips by Alkaline Condensation of Aminopropyltriethoxysilane

Directory of Open Access Journals (Sweden)

M. Ghorbanpour

2015-04-01

Full Text Available The Silver SPR chip was modified by alkaline-silane condensation with aminopropyltriethoxysilane (APTES in NaOH aqueous solution at different times. Silver sputtered slides coated with APTES were immersed in NaOH solution, enabling us to produce silver surfaces homogeneously covered with APTES. The surface properties of grafted APTES on sputtered silver surface as a occasion of time were studied using SPR analysis, AFM and contact angle measurement. The mechanical and chemical stability of samples was assayed by tape test and NaCl test. The answers show that hydrolysis and condensation of APTES are activated in alkaline solution and lead to formation of a protective APTES layer on the surface of silver. The morphology of APTES on silver surface is a function of coverage density that is altered by changing time. At short times (< 30 min, APTES molecules physically adsorb to the surface leads to weak protection. At higher condensation times, APTES molecules chemically bond to the surface and each other leads to better protection.

RB research nuclear reactor, Annual report for 1989, I - III

International Nuclear Information System (INIS)

Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

1989-12-01

This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

Near Infrared Characterization of Hetero-Core Optical Fiber SPR Sensors Coated with Ta2O5 Film and Their Applications

Directory of Open Access Journals (Sweden)

Kazuhiro Watanabe

2012-02-01

Full Text Available This paper describes the characteristics of optical fiber sensors with surface plasmon resonance (SPR at 1,310 nm in which the scattering loss of silica optical fiber is low. SPR operation in the infrared wavelength range is achieved by coating a thin tantalum pentaoxide (Ta2O5 film. The novelty of this paper lies in the verification of how the hetero-core scheme could be operated as a commercial base candidate in the sense of easy fabrication, sufficient mechanical strength, and significant sensitivity as a liquid detector under the basis of a low loss transmission network in the near infrared wavelength region. The effect of Ta2O5 layer thickness has been experimentally revealed in the wavelength region extending to 1,800 nm by using the hetero-core structured optical fiber. SPR characterizations have been made in the wavelength region 1,000–1,300 nm, showing the feasible operation at the near infrared wavelength and the possible practical applications. In addition, the technique developed in this work has been interestingly applied to a multi-point water-detection and a water-level gauge in which tandem-connected SPR sensors system using hetero-core structured fibers were incorporated. The detailed performance characteristics are also shown on these applications.

Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ

International Nuclear Information System (INIS)

Delfin L, A.; Palacios, J.C.; Alonso, G.

2006-01-01

Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

Application of fuel management calculation codes for CANDU reactor

International Nuclear Information System (INIS)

Ju Haitao; Wu Hongchun

2003-01-01

Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

Critical experiments at Sandia National Laboratories

International Nuclear Information System (INIS)

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-01-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-III is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark

Reflectance spectra characteristics from an SPR grating fabricated by nano-imprint lithography technique for biochemical nanosensor applications

Science.gov (United States)

Setiya Pradana, Jalu; Hidayat, Rahmat

2018-04-01

In this paper, we report our research work on developing a Surface Plasmon Resonance (SPR) element with sub-micron (hundreds of nanometers) periodicity grating structure. This grating structure was fabricated by using a simple nano-imprint lithography technique from an organically siloxane polymers, which was then covered by nanometer thin gold layer. The formed grating structure was a very well defined square-shaped periodic structure. The measured reflectance spectra indicate the SPR wave excitation on this grating structure. For comparison, the simulations of reflectance spectra have been also carried out by using Rigorous Coupled-Wave Analysis (RCWA) method. The experimental results are in very good agreement with the simulation results.

Comparison of different photoresist buffer layers in SPR sensors based on D-shaped POF and gold film

Science.gov (United States)

Cennamo, Nunzio; Pesavento, Maria; De Maria, Letizia; Galatus, Ramona; Mattiello, Francesco; Zeni, Luigi

2017-04-01

A comparative analysis of two optical fiber sensing platforms is presented. The sensors are based on surface plasmon resonance (SPR) in a D-shaped plastic optical fiber (POF) with a photoresist buffer layer between the exposed POF core and the thin gold film. We show how the sensor's performances change when the photoresist layer changes. The photoresist layers proposed in this analysis are SU-8 3005 and S1813. The experimental results are congruent with the numerical studies and it is instrumental for chemical and bio-chemical applications. Usually, the photoresist layer is required in order to increase the performance of the SPR-POF sensor.

Development of a β-Lactoglobulin Sensor Based on SPR for Milk Allergens Detection

DEFF Research Database (Denmark)

Ashley, Jon; D'Aurelio, Roberta; Piekarska, Monika

2018-01-01

A sensitive and label-free surface plasmon resonance (SPR) based sensor was developed in this work for the detection of milk allergens. β-lactoglobulin (BLG) protein was used as the biomarker for cow milk detection. This is to be used directly in final rinse samples of cleaning in-place (CIP) sys...

A SPR-based immunosensor for the detection of isoproturon.

Science.gov (United States)

Gouzy, Marie-Françoise; Kess, Melanie; Krämer, Petra M

2009-02-15

The proof of principle of a reusable surface plasmon resonance (SPR)-based immunosensor for the monitoring of isoproturon (IPU), a selective and systemic herbicide, is presented. The detecting rat monoclonal anti-isoproturon antibody (mAb IOC 7E1) was reversibly immobilized through the use of a capture mouse anti-rat (kappa-chain) monoclonal antibody (mAb TIB 172), which was covalently immobilized on the sensor chip surface. Such strategy features a controlled binding of the captured detecting antibody as well as facilitates the surface regeneration. The capture of the anti-IPU mAb by the antibody (TIB 172) coated sensor surface could be carried out up to 120 times (immobilization/regeneration cycles) without any evidence of activity loss. With a high test midpoint and a low associated SPR signal, the direct detection format was shown to be unsuitable for the routine analysis of isoproturon. However, the limit of detection (LOD) could be easily enhanced by using a strategy based on a surface competition assay, which improved all immunosensor parameters. Moreover, the sensitivity and working range of the indirect format were found to be dependent on the surface density of the anti-IPU mAb IOC 7E1. As expected for competitive formats, the lowest surface coverage (0.5 ng/mm(2)) allowed a lower detection of the herbicide isoproturon with a calculated LOD of 0.1 microg/l, an IC(50) (50% inhibition) of 5.3+/-0.6 microg/l, and a working range (20-80% inhibition) of 1.3-16.3 microg/l.

Inspection with non destructive assay techniques of the aluminium coating of the TRIGA Mark III reactor vat; Inspeccion con tecnicas de ensayos no destructivos del recubrimiento de aluminio de la tina del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Reyes A, A.I.; Gonzalez M, A.; Castaneda J, G.; Rivera M, H.; Sandoval G, I. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2001-07-01

In June 2000, the Reactor Department assigned to the Scientific Research Direction of the National Institute of Nuclear Research requested to the Non-destructive Assays Laboratory (LEND), assigned to the Materials Science Management, the inspection and measurement of thickness of the aluminium coating (liner) of the TRIGA Mark III reactor vat with non-destructive assay techniques, due to that the aluminium coating is exposed mainly to undergo slimming on its back side due to corrosion phenomena. Activity that was able to be carried out from april until august 2001. It is worth pointing out that this type of inspection with these techniques was realized by first time. The non-destructive assays (NDA) are techniques which use indirect physical methods for inspecting the sanitation of components in process or in service, for detect lack of continuity or defects which affect their quality or usefulness. The application of those do not alter the physical, chemical, mechanical or dimensional properties of the part subject of inspection. The results of the application of the ultrasound inspection techniques, industrial radiography and penetrating liquids are presented. (Author)

Phosphated in aluminium 6061-T651 used in the pool of the TRIGA Mark III nuclear reactor; Fosfatado en aluminio 6061-T651 utilizado en la tina del reactor nuclear TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F.; Espinosa L, J.; Pena B, A.; Perez F, C.; Sanchez C, M.; Vite T, M.; Vite T, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2001-07-01

We characterized a phosphated coating used a substrate in aluminium 6061-T651, which is used in the container of the TRIGA Mark III nuclear reactor. Characterization was made using MEB and X-ray diffraction techniques. Coating application has the function to prevent the corrosion. Coating was probed to test adhesion in accordance with the Standard ASTM D-4541, and the corrosion process using a salt spray (fog) camera, in accordance with the Standard ASTM B-117, so as we could phosphate the welding cord. These experiences were obtained using a Deep cell. Results obtained are going to phosphate 'in situ' using a mobile device which was patented for the National Institute of Nuclear Research (ININ) in the Mexican Institute of Intellectual Property (INPI). (Author)

Molecular self assembly of mixed comb-like dextran surfactant polymers for SPR virus detection.

Science.gov (United States)

Mai-Ngam, Katanchalee; Kiatpathomchai, Wansika; Arunrut, Narong; Sansatsadeekul, Jitlada

2014-11-04

The synthesis of two comb-like dextran surfactant polymers, that are different in their dextran molecular weight (MW) distribution and the presence of carboxylic groups, and their characterization are reported. A bimodal carboxylic dextran surfactant polymer consists of poly(vinyl amine) (PVAm) backbone with carboxyl higher MW dextran, non-functionalized lower MW dextran and hydrophobic hexyl branches; while a monomodal dextran surfactant polymer is PVAm grafted with non-functionalized lower MW dextran and hexyl branches. Layer formation of non-covalently attached dextran chains with bimodal MW distributions on a surface plasmon resonance (SPR) chip was investigated from the perspective of mixed physisorption of the bimodal and monomodal surfactant polymers. Separation distances between the carboxylic longer dextran side chains within the bimodal surfactant polymer and between the whole bimodal surfactant molecules on the chip surface could be well-controlled. SPR analysis of shrimp yellow head virus using our mixed surfactant chips showed dependence on synergetic adjustment of these separation distances. Copyright © 2014 Elsevier Ltd. All rights reserved.

Study of Immobilization Procedure on Silver Nanolayers and Detection of Estrone with Diverged Beam Surface Plasmon Resonance (SPR Imaging

Directory of Open Access Journals (Sweden)

Ibrahim Abdulhalim

2013-03-01

Full Text Available An immobilization protocol was developed to attach receptors on smooth silver thin films. Dense and packed 11-mercaptoundecanoic acid (11-MUA was used to avoid uncontrolled sulfidization and harmful oxidation of silver nanolayers. N,N'-dicyclohexylcarbodiimide (DCC and N-hydroxysuccinimide (NHS were added to make the silver surfaces reactive. A comparative study was carried out with different immersion times of silver samples in 11-MUA solutions with different concentrations to find the optimum conditions for immobilization. The signals, during each step of the protocol, were analyzed with a refractometer based on the surface plasmon resonance (SPR effect and luminescence techniques. Molecular interactions at the surfaces between the probe and target at the surface nanolayer shift the SPR signal, thus indicating the presence of the substance. To demonstrate specific biosensing, rabbit anti-estrone polyclonal immunoglobulin G (IgG antibody was immobilized through a linker on 47 nm silver layer deposited on SF11 glass. At the final stage, the representative endocrine disruptor—estrone—was attached and detected in deionized water with a diverging beam SPR imaging sensor.

Tuning the field distribution and fabrication of an Al@ZnO core-shell nanostructure for a SPR-based fiber optic phenyl hydrazine sensor.

Science.gov (United States)

Tabassum, Rana; Kaur, Parvinder; Gupta, Banshi D

2016-05-27

We report the fabrication and characterization of a surface plasmon resonance (SPR)-based fiber optic sensor that uses coatings of silver and aluminum (Al)-zinc oxide (ZnO) core-shell nanostructure (Al@ZnO) for the detection of phenyl hydrazine (Ph-Hyd). To optimize the volume fraction (f) of Al in ZnO and the thickness of the core-shell nanostructure layer (d), the electric field intensity along the normal to the multilayer system is simulated using the two-dimensional multilayer matrix method. The Al@ZnO core-shell nanostructure is prepared using the laser ablation technique. Various probes are fabricated with different values of f and an optimized thickness of core-shell nanostructure for the characterization of the Ph-Hyd sensor. The performance of the Ph-Hyd sensor is evaluated in terms of sensitivity. It is found that the Ag/Al@ZnO nanostructure core-shell-coated SPR probe with f = 0.25 and d = 0.040 μm possesses the maximum sensitivity towards Ph-Hyd. These results are in agreement with the simulated ones obtained using electric field intensity. In addition, the performance of the proposed probe is compared with that of probes coated with (i) Al@ZnO nanocomposite, (ii) Al nanoparticles and (iii) ZnO nanoparticles. It is found that the probe coated with an Al@ZnO core-shell nanostructure shows the largest resonance wavelength shift. The detailed mechanism of the sensing (involving chemical reactions) is presented. The sensor also manifests optimum performance at pH 7.

Energy deposition measurements in fast reactor safety experiments with fission thermocouple detectors

International Nuclear Information System (INIS)

Wright, S.A.; Scott, H.L.

1979-01-01

The investigation of phenomena occurring in in-pile fast reactor safety experiments requires an accurate measurement of the time dependent energy depositions within the fissile material. At Sandia Laboratories thin-film fission thermocouples are being developed for this purpose. These detectors have high temperature capabilities (400 to 500 0 C), are sodium compatible, and have milli-second time response. A significant advantage of these detectors for use as energy deposition monitors is that they produce an output voltage which is directly dependent on the temperature of a small chip of fissile material within the detectors. However, heat losses within the detector make it necessary to correct the response of the detector to determine the energy deposition. A method of correcting the detector response which uses an inverse convolution procedure has been developed and successfully tested with experimental data obtained in the Sandia Pulse Reactor (SPR-II) and in the Annular Core Research Reactor

Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment; Simulador de la cinetica puntual de un reactor nuclear TRIGA Mark III con control difuso de potencia en un ambiente visual

Energy Technology Data Exchange (ETDEWEB)

Perez M, C

2004-07-01

The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

Energy Technology Data Exchange (ETDEWEB)

Gallardo S, L F

1991-12-31

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

Nanoscale Affinity Chip Interface for Coupling Inhibition SPR Immunosensor Screening with Nano-LC TOF MS

NARCIS (Netherlands)

Marchesini, G.R.; Buijs, J.; Haasnoot, W.; Hooijerink, H.; Jansson, O.; Nielen, M.W.F.

2008-01-01

The on-line nanoscale coupling of a surface plasmon resonance (SPR)-based inhibition biosensor immunoassay (iBIA) for the screening of low molecular weight molecules with nano-liquid-chromatography electrospray ionization time-of-flight mass spectrometry (nano-LC ESI TOF MS) for identification is

Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun Sook; Lee, Byung Doo

2007-11-15

Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report.

Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

International Nuclear Information System (INIS)

Kim, Hyun Sook; Lee, Byung Doo

2007-11-01

Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report

SPR Characteristics Curve and Distribution of Residual Stress in Self-Piercing Riveted Joints of Steel Sheets

Directory of Open Access Journals (Sweden)

Rezwanul Haque

2017-01-01

Full Text Available Neutron diffraction was used to describe the residual stress distributions in self-piercing riveted (SPR joints. The sheet material displayed a compressive residual stress near the joint, and the stress gradually became tensile in the sheet material far away from the joint. The stress in the rivet leg was lower in the thick joint of the softer steel sheet than in the thin joint of the harder steel sheet. This lower magnitude was attributed to the lower force gradient during the rivet flaring stage of the SPR process curve. This study shows how the residual stress results may be related to the physical occurrences that happened during joining, using the characteristics curve. The study also shows that neutron diffraction technique enabled a crack in the rivet tip to be detected which was not apparent from a cross-section.

Physics of Fast and Intermediate Reactors. V. I. Proceedings of the Seminar on the Physics of Fast and Intermediate Reactors. V. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1962-03-15

It is generally agreed that the ultimate economic advantage of power produced by nuclear fission over that produced by conventional sources depends on the ability of a certain type of reactor to breed precious nuclear fuel out of the plentiful but not readily fissionable isotope of uranium. This fact is mainly responsible for the importance attached to the development of fast power reactors, but many other interesting properties of unmoderated or weakly moderated reactor systems have also been brought to light by reactor physicists. In August 1961 the Agency organized in Vienna a Seminar on the Physics of Past and Intermediate Reactors, at which all the topics relating to this important branch, of reactor science were discussed. The main feature of this meeting was extensive discussion of the 66 written contributions, which set the stage for a wide exchange of experience and ideas throughout 13 half-day sessions. The Seminar was attended by 132 scientists from 22 Member States and two international organizations. It is hoped that these Proceedings of the Seminar, which include both the papers presented and a record of the discussions, will be useful as a reference work both to research workers in the field and to newcomers to it for many years to come. The Agency's thanks are due to all the participating scientists for their written or oral contributions and especially to those among them who, as session chairmen, led the discussions and contributed greatly to the success of the meeting. During the Seminar, sixty-five papers were orally presented, and seven more were accepted for publication in the Proceedings. In order that these Proceedings might be in the hands of their users at an early date, the method of presentation of the papers and of the extensive session discussions had to be somewhat different from the one usually followed. The complete record of the sessions will be found at the end of Volume III. The order in which the papers are presented here is not

10 CFR 1049.8 - Training of SPR Protective Force Officers and qualification to carry firearms.

Science.gov (United States)

2010-01-01

... sufficient to maintain at least the minimum level of competency required for the successful performance of... competence to perform tasks associated with their responsibilities. The basic course shall include the...) Operating in such a manner as to preserve SPR sites and facilities; (9) Communications, including methods...

Critical experiments at Sandia National Laboratories

International Nuclear Information System (INIS)

Harms, G.A.; Ford, J.T.; Barber, A.D.

2011-01-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-III is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark

Critical experiments at Sandia National Laboratories

Energy Technology Data Exchange (ETDEWEB)

Harms, G.A.; Ford, J.T.; Barber, A.D., E-mail: gaharms@sandia.gov [Sandia National Laboratories, Albuquerque, NM (United States)

2011-07-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-III is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide

Historical Cavern Floor Rise for All SPR Sites

Energy Technology Data Exchange (ETDEWEB)

Moriarty, Dylan Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-09-01

The Strategic Petroleum Reserve (SPR) contains the largest supply is the largest stockpile of government-owned emergency crude oil in the world. The oil is stored in multiple salt caverns spread over four sites in Louisiana and Texas. Cavern infrastructure near the bottom of the cavern can be damaged from vertical floor movement. This report presents a comprehensive history of floor movements in each cavern. Most of the cavern floor rise rates ranged from 0.5-3.5 ft/yr, however, there were several caverns with much higher rise rates. BH103, BM106, and BH105 had the three highest rise rates. Information from this report will be used to better predict future vertical floor movements and optimally place cavern infrastructure. The reasons for floor rise are not entirely understood and should be investigated.

Issues of high-burnup fuel for advanced nuclear reactors

International Nuclear Information System (INIS)

Belac, J.; Milisdoerfer, L.

2004-12-01

A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

Comparison between a Direct-Flow SPR Immunosensor for Ampicillin and a Competitive Conventional Amperometric Device: Analytical Features and Possible Applications to Real Samples

Science.gov (United States)

Tomassetti, Mauro; Merola, Giovanni; Martini, Elisabetta; Campanella, Luigi; Sanzò, Gabriella; Favero, Gabriele; Mazzei, Franco

2017-01-01

In this research, we developed a direct-flow surface plasmon resonance (SPR) immunosensor for ampicillin to perform direct, simple, and fast measurements of this important antibiotic. In order to better evaluate the performance, it was compared with a conventional amperometric immunosensor, working with a competitive format with the aim of finding out experimental real advantages and disadvantages of two respective methods. Results showed that certain analytical features of the new SPR immunodevice, such as the lower limit of detection (LOD) value and the width of the linear range, are poorer than those of a conventional amperometric immunosensor, which adversely affects the application to samples such as natural waters. On the other hand, the SPR immunosensor was more selective to ampicillin, and measurements were more easily and quickly attained compared to those performed with the conventional competitive immunosensor. PMID:28394296

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

Energy Technology Data Exchange (ETDEWEB)

Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

2008-08-15

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico; Evaluacion de la aptitud para el servicio de la piscina del reactor TRIGA Mark III del Instituto Nacional de Investigaciones Nucleares de Mexico

Energy Technology Data Exchange (ETDEWEB)

Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J. [Departamento de Sintesis y Caracterizacion de Materiales del ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2001-07-01

This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

Inspection with non destructive assay techniques of the aluminium coating of the TRIGA Mark III reactor vat

International Nuclear Information System (INIS)

Reyes A, A.I.; Gonzalez M, A.; Castaneda J, G.; Rivera M, H.; Sandoval G, I.

2001-01-01

In June 2000, the Reactor Department assigned to the Scientific Research Direction of the National Institute of Nuclear Research requested to the Non-destructive Assays Laboratory (LEND), assigned to the Materials Science Management, the inspection and measurement of thickness of the aluminium coating (liner) of the TRIGA Mark III reactor vat with non-destructive assay techniques, due to that the aluminium coating is exposed mainly to undergo slimming on its back side due to corrosion phenomena. Activity that was able to be carried out from april until august 2001. It is worth pointing out that this type of inspection with these techniques was realized by first time. The non-destructive assays (NDA) are techniques which use indirect physical methods for inspecting the sanitation of components in process or in service, for detect lack of continuity or defects which affect their quality or usefulness. The application of those do not alter the physical, chemical, mechanical or dimensional properties of the part subject of inspection. The results of the application of the ultrasound inspection techniques, industrial radiography and penetrating liquids are presented. (Author)

CCF analysis of BWR reactor shutdown systems based on the operating experience at the TVO I/II in 1981-1993

International Nuclear Information System (INIS)

Mankamo, T.

1996-04-01

The work constitutes a part of the project conducted within the research program of the Swedish Nuclear Power Inspectorate SKI, aimed to develop the methods and data base for the Common Cause Failure (CCF) analysis of highly redundant reactor scram systems. The data analysis for the TVO I/II plant is focused on the hydraulic scram system, and control rods and drives. It covers operating experiences from 1981 through 1993. (9 refs., 9 figs., 7 tabs.)

Energy conversion options for ARIES-III - A conceptual D-3He tokamak reactor

International Nuclear Information System (INIS)

Santarius, J.F.; Blanchard, J.P.; Emmert, G.A.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Ghoneim, N.M.; Hasan, M.Z.; Mau, T.K.; Greenspan, E.; Herring, J.S.; Kernbichler, W.; Klein, A.C.; Miley, G.H.; Miller, R.L.; Peng, Y.K.M.

1989-01-01

The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D- 3 He fusion reactors. This stems from: (1) the large fraction of D- 3 He power produced in the forms of charged particles and synchrotron radiation which are amenable to direct conversion, and (2) the low neutron fluence and lack of tritium breeding constraints, which increase design flexibility. The design team for a conceptual D- 3 He tokamak reactor, ARIES-III, has investigated numerous energy conversion options at a scoping level in attempting to realize high efficiency. The energy conversion systems have been studied in the context of their use on one or more of three versions of a D- 3 He tokamak: a first stability regime device, a second stability regime device, and a spherical torus. The set of energy conversion options investigated includes bootstrap current conversion, compression-expansion cycles, direct electrodynamic conversion, electrostatic direct conversion, internal electric generator, liquid metal heat engine blanket, liquid metal MHD, plasma MHD, radiation boiler, scrape-off layer thermoelectric, synchrotron radiation conversion by rectennas, synchrotron radiation conversion by thermal cycles, thermionic/AMTEC/thermal systems, and traveling wave conversion. The original set of options is briefly discussed, and those selected for further study are described in more detail. The four selected are liquid metal MHD, plasma MHD, rectenna conversion, and direct electrodynamic conversion. Thermionic energy conversion is being considered, and some options may require a thermal cycle in parallel or series. 17 refs., 3 figs., 1 tab

Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Parra M, M. A.

2014-07-01

This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au{sup 197}, Ni{sup 58}, In{sup 115}, Mg{sup 24}, Al{sup 27}, Fe{sup 58}, Co{sup 59} and Cu{sup 63}, they were selected to cover the full range the energies (10{sup -10} to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or

Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core; Programas de captura, analisis y graficado de datos para el estudio de la termometria del nucleo del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Paredes G, L.C

1991-05-15

This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core

International Nuclear Information System (INIS)

Paredes G, L.C.

1991-05-01

This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

Immobilization of flavan-3-ols onto sensor chips to study their interactions with proteins and pectins by SPR

Energy Technology Data Exchange (ETDEWEB)

Watrelot, Aude A., E-mail: aude.watrelot@avignon.inra.fr [INRA, UMR408 Sécurité et Qualité des Produits d’Origine Végétale, Domaine St Paul, Site Agroparc, 84914 Avignon (France); Université d’Avignon, UMR408 Sécurité et Qualité des Produits d' Origine Végétale, F-84000 Avignon (France); Tran, Dong Tien [Université de Bordeaux, Institut des Sciences Moléculaires (UMR-CNRS 5255), 351 cours de la Libération, 33405 Talence (France); Institut Européen de Chimie et Biologie (IECB), 2 rue Robert Escarpit, 33607 Pessac (France); Buffeteau, Thierry [Université de Bordeaux, Institut des Sciences Moléculaires (UMR-CNRS 5255), 351 cours de la Libération, 33405 Talence (France); Deffieux, Denis [Université de Bordeaux, Institut des Sciences Moléculaires (UMR-CNRS 5255), 351 cours de la Libération, 33405 Talence (France); Institut Européen de Chimie et Biologie (IECB), 2 rue Robert Escarpit, 33607 Pessac (France); and others

2016-05-15

Highlights: • Flavanol-macromolecule interactions were determined using SPR. • Flavanols were chemically modified with a linker bearing a thiol group. • Flavanols were immobilized onto a carboxymethyl dextran surface. • Citrus pectin interacted more with flavanols than apple pectin. • Epicatechin interacted more with BSA than flavanol oligomer. - Abstract: Interactions between plant polyphenols and biomacromolecules such as proteins and pectins have been studied by several methods in solution (e.g. isothermal titration calorimetry, dynamic light scattering, nuclear magnetic resonance and spectrophotometry). Herein, these interactions were investigated in real time by Surface Plasmon Resonance (SPR) analysis after immobilization of flavan-3-ols onto a sensor chip surface. (−)-epicatechin, (+)-catechin and flavan-3-ol oligomers with an average degree of polymerization of 2 and 8 were chemically modified using N-(2-(tritylthio)ethyl)propiolamide in order to introduce a spacer unit onto the catecholic B ring. Modified flavan-3-ols were then immobilized onto a carboxymethylated dextran surface (CM5). Immobilization was validated and further verified by evaluating flavan-3-ol interaction with bovine serum albumin (BSA), poly-L-proline or commercial pectins. BSA was found to have a stronger association with monomeric flavan-3-ols than oligomers. SPR analysis of selected flavan-3-ols immobilized onto CM5 sensor chips showed a stronger association for citrus pectins than apple pectins, regardless of flavan-3-ol degree of polymerization.

Immobilization of flavan-3-ols onto sensor chips to study their interactions with proteins and pectins by SPR

International Nuclear Information System (INIS)

Watrelot, Aude A.; Tran, Dong Tien; Buffeteau, Thierry; Deffieux, Denis

2016-01-01

Highlights: • Flavanol-macromolecule interactions were determined using SPR. • Flavanols were chemically modified with a linker bearing a thiol group. • Flavanols were immobilized onto a carboxymethyl dextran surface. • Citrus pectin interacted more with flavanols than apple pectin. • Epicatechin interacted more with BSA than flavanol oligomer. - Abstract: Interactions between plant polyphenols and biomacromolecules such as proteins and pectins have been studied by several methods in solution (e.g. isothermal titration calorimetry, dynamic light scattering, nuclear magnetic resonance and spectrophotometry). Herein, these interactions were investigated in real time by Surface Plasmon Resonance (SPR) analysis after immobilization of flavan-3-ols onto a sensor chip surface. (−)-epicatechin, (+)-catechin and flavan-3-ol oligomers with an average degree of polymerization of 2 and 8 were chemically modified using N-(2-(tritylthio)ethyl)propiolamide in order to introduce a spacer unit onto the catecholic B ring. Modified flavan-3-ols were then immobilized onto a carboxymethylated dextran surface (CM5). Immobilization was validated and further verified by evaluating flavan-3-ol interaction with bovine serum albumin (BSA), poly-L-proline or commercial pectins. BSA was found to have a stronger association with monomeric flavan-3-ols than oligomers. SPR analysis of selected flavan-3-ols immobilized onto CM5 sensor chips showed a stronger association for citrus pectins than apple pectins, regardless of flavan-3-ol degree of polymerization.

An Advanced Semimetal-Organic Bi Spheres-g-C3N4 Nanohybrid with SPR-Enhanced Visible-Light Photocatalytic Performance for NO Purification.

Science.gov (United States)

Dong, Fan; Zhao, Zaiwang; Sun, Yanjuan; Zhang, Yuxin; Yan, Shuai; Wu, Zhongbiao

2015-10-20

To achieve efficient photocatalytic air purification, we constructed an advanced semimetal-organic Bi spheres-g-C3N4 nanohybrid through the in-situ growth of Bi nanospheres on g-C3N4 nanosheets. This Bi-g-C3N4 compound exhibited an exceptionally high and stable visible-light photocatalytic performance for NO removal due to the surface plasmon resonance (SPR) endowed by Bi metal. The SPR property of Bi could conspicuously enhance the visible-light harvesting and the charge separation. The electromagnetic field distribution of Bi spheres involving SPR effect was simulated and reaches its maximum in close proximity to the Bi particle surface. When the Bi metal content was controlled at 25%, the corresponding Bi-g-C3N4 displayed outstanding photocatalytic capability and transcended those of other visible-light photocatalysts. The Bi-g-C3N4 exhibited a high structural stability under repeated photocatalytic runs. A new visible-light-induced SPR-based photocatalysis mechanism with Bi-g-C3N4 was proposed on the basis of the DMPO-ESR spin-trapping. The photoinduced electrons could transfer from g-C3N4 to the Bi metal, as revealed with time-resolved fluorescence spectra. The function of Bi semimetal as a plasmonic cocatalyst for boosting visible light photocatalysis was similar to that of noble metals, which demonstrated a great potential of utilizing the economically feasible Bi element as a substitute for noble metals for the advancement of photocatalysis efficiency.

Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment

International Nuclear Information System (INIS)

Perez M, C.

2004-01-01

The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

Determination of the flows profile in the role of power in the central thimble of TRIGA Mark III Reactor; Determinacion del perfil de flujos en funcion de la potencia en el dedal central del Reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia F, A.

2010-07-01

The overall objective of the thesis project is to determine the flow profiles sub cadmic and epi cadmic in the central thimble to different powers and operation times of TRIGA Mark III Reactor, using activation foils as detectors. In the reactor operation, it is necessary to know the neutron flow profile for to realize other tasks as: the radioisotopes production, research in reactors physics and fuel burning. The distribution of the neutron flow, accurately reflects what is happening in the reactor core, plus the flows value in this distribution is directly related to the power generated. For this reason it is performed the sub cadmic flow measurement with energies between 0 and 0.4 eV (energy of the cadmium cut E{sub cd} approx 0.4 eV) and epi cadmic flow with energies greater than 0.4 eV, in the central thimble powers to the powers of 10, 100 W, 1, 10 100 Kw and 1 MW. The method used is known as flakes activation, which is to be arranged by placing flakes ( 3 mm of diameter and 0.0508 mm of thickness) of a given material (either Au, In, Cu, Mn, etc.) into an aluminum tube outside diameter equal to 6.35 mm, alternating flakes with lids covered and discovered of cadmium (3.4 mm of diameter and 0.508 mm of thickness) and separated by lucite pieces of 3 mm of diameter and 25.4 mm in length. After irradiating the flakes for some time, is measured the gamma activity of each of them, using a hyper pure germanium detector of high resolution. Already known gamma activity, proceed to calculate the epi cadmic and sub cadmic flows using a computer program in Fortran language, called Caflu. (Author)

Development of a β-Lactoglobulin Sensor Based on SPR for Milk Allergens Detection

Directory of Open Access Journals (Sweden)

Jon Ashley

2018-03-01

Full Text Available A sensitive and label-free surface plasmon resonance (SPR based sensor was developed in this work for the detection of milk allergens. β-lactoglobulin (BLG protein was used as the biomarker for cow milk detection. This is to be used directly in final rinse samples of cleaning in-place (CIP systems of food manufacturers. The affinity assay was optimised and characterised before a standard curve was performed in pure buffer conditions, giving a detection limit of 0.164 µg mL−1 as a direct binding assay. The detection limit can be further enhanced through the use of a sandwich assay and amplification with nanomaterials. However, this was not required here, as the detection limit achieved exceeded the required allergen detection levels of 2 µg mL−1 for β-lactoglobulin. The binding affinities of the polyclonal antibody for BLG, expressed by the dissociation constant (KD, were equal to 2.59 × 10−9 M. The developed SPR-based sensor offers several advantages in terms of label-free detection, real-time measurements, potential on-line system and superior sensitivity when compared to ELISA-based techniques. The method is novel for this application and could be applied to wider food allergen risk management decision(s in food manufacturing.

Application of the reactor kinetics equations to the reactor safety analysis

International Nuclear Information System (INIS)

Sdouz, G.

1976-01-01

The reactor kinetics equations which can be solved by the computer program AIREK-III are used to describe the behavior of fast reactivity transients. By supplementing this computer program it was possible to solve additional safety problems, e.g. the course of reactor excursions induced by any form of reactivity input, the control of reactivity input as a function of a threshold-energy and the computation of produced energy. (author)

Grating-Coupled Surface Plasmon Resonance (GC-SPR) Optimization for Phase-Interrogation Biosensing in a Microfluidic Chamber.

Science.gov (United States)

Rossi, Stefano; Gazzola, Enrico; Capaldo, Pietro; Borile, Giulia; Romanato, Filippo

2018-05-18

Surface Plasmon Resonance (SPR)-based sensors have the advantage of being label-free, enzyme-free and real-time. However, their spreading in multidisciplinary research is still mostly limited to prism-coupled devices. Plasmonic gratings, combined with a simple and cost-effective instrumentation, have been poorly developed compared to prism-coupled system mainly due to their lower sensitivity. Here we describe the optimization and signal enhancement of a sensing platform based on phase-interrogation method, which entails the exploitation of a nanostructured sensor. This technique is particularly suitable for integration of the plasmonic sensor in a lab-on-a-chip platform and can be used in a microfluidic chamber to ease the sensing procedures and limit the injected volume. The careful optimization of most suitable experimental parameters by numerical simulations leads to a 30⁻50% enhancement of SPR response, opening new possibilities for applications in the biomedical research field while maintaining the ease and versatility of the configuration.

Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Parra M, M. A.; Luis L, M. A. [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Division de Ciencias Basicas, Av. San Pablo No. 180, Col. Reynosa Tamaulipas, 02200 Mexico D. F. (Mexico); Raya A, R.; Cruz G, H. S., E-mail: roberto.raya@inin.gob.mx [ININ, Departamento del Reactor, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10{sup -10} to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of {sup 197}Au, {sup 58}Ni, {sup 115}In, {sup 24}Mg, {sup 27}Al, {sup 58}Fe, {sup 59}Co and {sup 63}Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

Upgrading program of the experimental fast reactor Joyo

International Nuclear Information System (INIS)

Yoshida, A.; Yogo, S.

2001-01-01

The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

Preparations for decommissioning the TRIGA Mark III Berkeley Research Reactor

International Nuclear Information System (INIS)

Denton, Michael M.; Lim, Tek. H.

1988-01-01

On December 20, 1986 the chancellor of UC Berkeley announced his decision to decommission the 20 year old Berkeley Research Reactor citing as principal reasons a decline in use and a need to erect a new computer science building over the reactor's site. In order to meet the University's construction timetable for the new building, the reactor staff together with other units of the campus administration have initiated a program to remove the reactor structure and clear the room for unlicensed use as expediently as possible. Due to the sequence of events which must occur in a limited amount of time, the University adopted a policy to contract out as much of the work as possible, including generation of the defueling and decommissioning plans.The first physical step in the decommissioning project is the removal of the irradiated fuel. This task is largely contracted out to a commercial firm with experience in the transport of radioactive materials and reactor fuel. As suggested by the NRC, the reactor will be defueled under the current operating license. This requires that all fuel must be off-site before the DP can be approved. Therefore any delay in defueling in-turn delays the decommissioning. The NRC has given no commitment or date for completion of their review. Informal discussion with NRC project managers and the experience from other facilities indicate that the review process will take between six and nine months

A wide-range medition system for TRIGA Mark III Reactor

International Nuclear Information System (INIS)

Vazquez R, M.

1995-01-01

The number of particles emitted by a fission reaction is proportional to the number of fissions generated in the reactor nucleus, what in turn are proportional to the power level of such reactor; this indicates that it is possible could measure the reactor power if the amount of neutrons is measured, or the density of the neutron flux. The power measurement in the reactor is necessary in order to have a control of the same. Several procedures for power measurement exists, one of them is achieved through neutron flux density measurement, that take place in the chain reaction by means of the flow density measurement. The easiest way in order to achieve the neutron flux density measurement is carry out by means of the employment of neutron detectors. To the exit of these detectors an electric pulse taks place every time that a neutron interacts with the sensitive detector part. The work here presented, concrete to the construction of a system of measurement of the nuclear power reactor, that is based on the neutron flux applying some techniques of the neutron noise analysis. (Author)

Development of low cost and accurate homemade sensor system based on Surface Plasmon Resonance (SPR)

Science.gov (United States)

Laksono, F. D.; Supardianningsih; Arifin, M.; Abraha, K.

2018-04-01

In this paper, we developed homemade and computerized sensor system based on Surface Plasmon Resonance (SPR). The developed systems consist of mechanical system instrument, laser power sensor, and user interface. The mechanical system development that uses anti-backlash gear design was successfully able to enhance the angular resolution angle of incidence laser up to 0.01°. In this system, the laser detector acquisition system and stepper motor controller utilizing Arduino Uno which is easy to program, flexible, and low cost, was used. Furthermore, we employed LabView’s user interface as the virtual instrument for facilitating the sample measurement and for transforming the data recording directly into the digital form. The test results using gold-deposited half-cylinder prism showed the Total Internal Reflection (TIR) angle of 41,34°± 0,01° and SPR angle of 44,20°± 0,01°, respectively. The result demonstrated that the developed system managed to reduce the measurement duration and data recording errors caused by human error. Also, the test results also concluded that the system’s measurement is repeatable and accurate.

Determination of the subcadmium flux in the Triga Mark III reactor of the Nuclear Center in Mexico

International Nuclear Information System (INIS)

Rodiguez V, F.J.

1975-01-01

The determination of the subcadmium flux profile (neutrons with energy between 0 and 0.4 eV approximatley) in the core of the reactor Triga Mark III in the Nuclear Center of Mexico was made. The technique used for that purpose consists in placing alternativelly indium sheets covered and uncovered with cadmium and separated by fragments of aluminium 2.54cm wide in an aluminium tube whose external diameter reaches 0.635cm. This tube is introduced in each of the 21 axial irradiation places of the sheets the activity of gamma rays of 1.293 MeV of In-116m produced during the irradiation, is measured using a monochannel analyzer. We obtain graphs of the specific counting ratio vs position for the sheets uncovered and covered with cadmium; using quadratic interpolation we compute in each position the missing specific measuring ratio. The difference between the specific counting ratios with our cadmium and with cadmium multiplied by a constant gives the subcadmium flux in the z point. For irradiation at 10 watts, 1 Kilowatt and 1 megawatt the results show profiles which in general terms are in accordance with those which were obtained theoretically for a cylyndrical reactor using group theory. The greatest flow given by the manufacturer of the reactor is in the values range obtained experimentally.For the calculations, the method uses a FORTRAN IV program so that the determinations can be made routinely. (author)

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

Fast reactor physics - an overview

International Nuclear Information System (INIS)

Lee, S.M.

2004-01-01

An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

Grating coupled SPR microarray analysis of proteins and cells in blood from mice with breast cancer.

Science.gov (United States)

Mendoza, A; Torrisi, D M; Sell, S; Cady, N C; Lawrence, D A

2016-01-21

Biomarker discovery for early disease diagnosis is highly important. Of late, much effort has been made to analyze complex biological fluids in an effort to develop new markers specific for different cancer types. Recent advancements in label-free technologies such as surface plasmon resonance (SPR)-based biosensors have shown promise as a diagnostic tool since there is no need for labeling or separation of cells. Furthermore, SPR can provide rapid, real-time detection of antigens from biological samples since SPR is highly sensitive to changes in surface-associated molecular and cellular interactions. Herein, we report a lab-on-a-chip microarray biosensor that utilizes grating-coupled surface plasmon resonance (GCSPR) and grating-coupled surface plasmon coupled fluorescence (GCSPCF) imaging to detect circulating tumor cells (CTCs) from a mouse model (FVB-MMTV-PyVT). GCSPR and GCSPCF analysis was accomplished by spotting antibodies to surface cell markers, cytokines and stress proteins on a nanofabricated GCSPR microchip and screening blood samples from FVB control mice or FVB-MMTV-PyVT mice with developing mammary carcinomas. A transgenic MMTV-PyVT mouse derived cancer cell line was also analyzed. The analyses indicated that CD24, CD44, CD326, CD133 and CD49b were expressed in both cell lines and in blood from MMTV-PyVT mice. Furthermore, cytokines such as IL-6, IL-10 and TNF-α, along with heat shock proteins HSP60, HSP27, HSc70(HSP73), HSP90 total, HSP70/HSc70, HSP90, HSP70, HSP90 alpha, phosphotyrosine and HSF-1 were overexpressed in MMTV-PyVT mice.

Estimation development cost, study case: Quality Management System Reactor TRIGA Mark III

International Nuclear Information System (INIS)

Antúnez Barbosa, Tereso Antonio; Valdovinos Rosas, Rosa María; Marcial Romero, José Raymundo; Ramos Corchado, Marco Antonio; Edgar Herrera Arriaga

2016-01-01

The process of estimating costs in software engineering is not a simple task, it must be addressed carefully to obtain an efficient strategy to solve problems associated with the effort, cost and time of activities that are performed in the development of an information system project. In this context the main goal for both developers and customers is the cost, since developers are worry about the effort pay-load and customers are worry about the product pay-load. However, in other fields the cost of goods depends on the activity or process that is performed, thereby deduce that the main cost of the final product of a development project software project is undoubtedly its size. In this paper a comparative study of common models for estimating costs are developed. These models are used today in order to create a structured analysis to provide the necessary information about cost, time and effort for making decisions in a software development project. Finally the models are applied to a case study, which is a system called Monitorizacion Automatica del Sistema de Gestion de Calidad del Reactor TRIGA Mark III. (author)

Theoretical evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core; Evaluacion teorica de la produccion de los venenos Xe-135 y Sm-149 del reactor TRIGA Mark III con nucleo mixto

Energy Technology Data Exchange (ETDEWEB)

Paredes G, L.C

1991-11-15

It was theoretically determined the accumulation of the Xe{sup 135} and Sm {sup 149} in function, of the time during a stationary state of 72 h. continuous for the reactor TRIGA Mark III to 1 MW of thermal power with mixed core. The values of negative reactivity due to these isotopes are of 2.04 dollars and 0.694 dollars to the 72 h, quantities that will have to be compensated if wants that the reactor continues working to this power. Under the same conditions but considering a core with standard fuel, it was found a value of {rho} = 1.70 dollars, resulting a difference of 0.30 dollars of negative reactivity in function of the type of analyzed core. This difference is important for the calculations of fuel management of a reactor. The concentration in balance of the xenon was reaches after an operation to constant power of 1 MW by 50 h, contrary to the samarium that reaches it balance after 3 weeks of operation starting from the initial start up and it stays constant along the useful life of the reactor while a change of fuel doesn't exist. It was obtained that for operation times greater to 60 h. at 1 MW, a peak of negative reactivity of the Xe{sup 135} is generated between the 7 and 11 h after the instantaneous shut down, with a value of 2.43 dollars, that is to say 0.39 additional dollars to those taken place during the continuous irradiation. (Author)

Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

2012-10-01

Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ; Diseno asistido por computadora (DAC) para mejorar la electronica de los canales nucleares del reactor TRIGA Mark III del ININ

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jlgm@nuclear.inin.mx

2007-07-01

The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

Multiplexed imaging surface plasmon resonance (iSPR) biosensor assay for the detection of Fusarium toxins in wheat

Science.gov (United States)

Certain Fusarium species (F. graminearum and F. verticilloides in particular) infest grains and can produce a wide range of fungal (myco)-toxins, causing huge economic losses worldwide. A reproducible and sensitive imaging surface plasmon resonance (iSPR) assay was developed and validated for three ...

Target immobilization as a strategy for NMR-based fragment screening: comparison of TINS, STD, and SPR for fragment hit identification.

Science.gov (United States)

Kobayashi, Masakazu; Retra, Kim; Figaroa, Francis; Hollander, Johan G; Ab, Eiso; Heetebrij, Robert J; Irth, Hubertus; Siegal, Gregg

2010-09-01

Fragment-based drug discovery (FBDD) has become a widely accepted tool that is complementary to high-throughput screening (HTS) in developing small-molecule inhibitors of pharmaceutical targets. Because a fragment campaign can only be as successful as the hit matter found, it is critical that the first stage of the process be optimized. Here the authors compare the 3 most commonly used methods for hit discovery in FBDD: high concentration screening (HCS), solution ligand-observed nuclear magnetic resonance (NMR), and surface plasmon resonance (SPR). They selected the commonly used saturation transfer difference (STD) NMR spectroscopy and the proprietary target immobilized NMR screening (TINS) as representative of the array of possible NMR methods. Using a target typical of FBDD campaigns, the authors find that HCS and TINS are the most sensitive to weak interactions. They also find a good correlation between TINS and STD for tighter binding ligands, but the ability of STD to detect ligands with affinity weaker than 1 mM K(D) is limited. Similarly, they find that SPR detection is most suited to ligands that bind with K(D) better than 1 mM. However, the good correlation between SPR and potency in a bioassay makes this a good method for hit validation and characterization studies.

Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident

International Nuclear Information System (INIS)

Raisic, N.

1963-02-01

Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects

Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III

International Nuclear Information System (INIS)

Parra M, M. A.

2014-01-01

This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au 197 , Ni 58 , In 115 , Mg 24 , Al 27 , Fe 58 , Co 59 and Cu 63 , they were selected to cover the full range the energies (10 -10 to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or research projects. (Author)

Artificial intelligence in nuclear reactor operation

International Nuclear Information System (INIS)

Da Ruan; Benitez-Read, J.S.

2005-01-01

Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

Cloning, expression and structural stability of a cold-adapted ß-Galactosidase from Rahnella sp.R3

Science.gov (United States)

A novel gene was isolated for the first time from a psychrophilic gram-negative bacterium Rahnella sp.R3. It encoded a cold-adapted ß-galactosidase (R-ß-Gal). Recombinant R-ß-Gal was expressed in Escherichia coli BL21 (DE3), purified, and characterized. R-ß-Gal belongs to the glycosyl hydrolase fami...

Status and some safety philosophies of the China advanced research reactor CARR

International Nuclear Information System (INIS)

Luzheng Yuan

2001-01-01

The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21 st century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D 2 O reflector not less than 8 x 10 14 n/cm 2 . s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10 15 n/cm 2 . s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

SFG and SPR Study of Sodium Dodecyl Sulfate Film Assembly on Positively Charged Surfaces

Science.gov (United States)

Song, Sanghun; Weidner, Tobias; Wagner, Matthew; Castner, David

2012-02-01

This study uses sum frequency generation (SFG) vibrational spectroscopy and surface plasmon resonance (SPR) sensing to investigate the structure of sodium dodecyl sulfate (SDS) films formed on positively charged and hydrophilic surfaces. The SPR signals show a good surface coverage suggesting that full monolayer coverage is reached at 1 mM. SFG spectra of SDS adsorbed exhibits well resolved CH3 peaks and OH peaks. At both 0.2 mM and 1 mM SDS concentration the intensity of both the CH3 and OH peaks decreased close to background levels. We found that the loss of SFG signal at 0.2 mM occurs at this concentration independent of surface charge density. It is more likely that the loss of signal is related to structural inhomogeneity induced by a striped phase - stand-up phase transition. This is supported by a distinct change of the relative SFG phase between CH3/OH near 0.2 mM. The second intensity minimum might be related to charge compensation effects. We observed a substrate dependence for the high concentration transition. We also observed distinct SFG signal phase changes for water molecules associated with SDS layers at different SDS solution concentrations indicating that the orientation of bound water changed with SDS surface structure.

Gold nanoparticles decorated on BaTiO3 as photocatalyst: effect of SPR and ferroelectricity

Science.gov (United States)

Miao, Zhilei; Chen, Lei; Wang, Rui; Yuan, Rongchun; Zhou, Fang; Lv, Pengfei; Zhang, Xiuyun; Wang, Qiang

2018-02-01

BaTiO3(BTO) powders with cubic(C) and tetragonal(T) crystalline structures were prepared by sol-gel method followed by calcination at different temperatures, and they were decorated with gold nanoparticles(AuNPs) to form Au-T-BTO and Au-C-BTO respectively. All the samples (C-BTO, T-BTO, Au-C-BTO and Au-T-BTO) were used as photo-catalysts for the degradation of Rhodamine B (RhB). Au-T-BTO exhibits the highest photocatalytic activity due to thecombined effect of surface plasmon resonance (SPR) and the ferroelectricity of the tetragonalphase BTO. The light absorption peaks at 500 to 600 nm verify the presence of SPR effect from Au NPs. Based on density functional theory (DFC) within the generalized gradient approximation (GGA) approach, it was demonstrated that the tetragonal phase BaTiO3 shows a spontaneous polarization with the calculated value of 0.34 C m-2, which is absent in thecubic phase. The internal space charge layer in tetragonal phase BTO enhances the separation of photoexcited carriers due to the spontaneous ferroelectric polarization, which also benefits photocatalytic activities.

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

Phosphated in aluminium 6061-T651 used in the pool of the TRIGA Mark III nuclear reactor

International Nuclear Information System (INIS)

Aguilar H, F.; Espinosa L, J.; Pena B, A.; Perez F, C.; Sanchez C, M.; Vite T, M.; Vite T, J.

2001-01-01

We characterized a phosphated coating used a substrate in aluminium 6061-T651, which is used in the container of the TRIGA Mark III nuclear reactor. Characterization was made using MEB and X-ray diffraction techniques. Coating application has the function to prevent the corrosion. Coating was probed to test adhesion in accordance with the Standard ASTM D-4541, and the corrosion process using a salt spray (fog) camera, in accordance with the Standard ASTM B-117, so as we could phosphate the welding cord. These experiences were obtained using a Deep cell. Results obtained are going to phosphate 'in situ' using a mobile device which was patented for the National Institute of Nuclear Research (ININ) in the Mexican Institute of Intellectual Property (INPI). (Author)

Conversion of the core of the TRIGA Mark III reactor at the Mexican Nuclear Centre

International Nuclear Information System (INIS)

Moran Lopez, J.M.; Lucatero, M.A.; Reyes Andrade, B.; Rivero Gutierrez, T.; Sainz Mejia, E.

1990-01-01

It was decided to convert the core of the TRIGA MARK III reactor at the Mexican Nuclear Centre run by the National Nuclear Institute because of problems detected during the operation, such as a lack of excess reactivity for operation at nominal power over long periods and difficulties in the maintenance and calibration of the control panel. In order to compensate for the lack of excess reactivity the fuel elements taken to the highest burnup were replaced by fresh elements acquired for this purpose. The latter, however, had a different enrichment, and this necessitated a detailed analysis of the neutronic and thermohydraulic behaviour of the reactor with a view to determining a mixed core configuration which would meet safe operation requirements. In conducting the thermohydraulic analysis, a natural convection coolant flow model was developed to determine coolant velocity and pressure drop patterns within the core. The heat transfer equations were solved and it was found that the hottest fuel element did not attain critical heat flux conditions. In loading this core it was also necessary to analyse procedures and to consider the possible effects of reaching criticality with fuel elements having different enrichments. The loading procedure is described, as is the measurement system and the results obtained. In order to resolve the calibration and maintenance problems, a new, more advanced control panel was designed with conventional and nuclear detection systems and modern components

RB research nuclear reactor, Annual report for 1984, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1984. godini, I - III

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1984-07-01

The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source.

An Axiomatic Design Approach of Nanofluid-Engineered Nuclear Safety Features for Generation III+ React

International Nuclear Information System (INIS)

Bang, In Cheol; Heo, Gyun Young; Jeong, Yong Hoon; Heo, Sun

2009-01-01

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems

Synthesis of surface plasmon resonance (SPR) triggered Ag/TiO2 photocatalyst for degradation of endocrine disturbing compounds

International Nuclear Information System (INIS)

Leong, Kah Hon; Gan, Bee Ling; Ibrahim, Shaliza; Saravanan, Pichiah

2014-01-01

Graphical abstract: - Highlights: • Ag/TiO 2 was synthesized with aid of natural photon stimulated photoreduction. • Deposited Ag prompted well the LSPRs, Schottky barrier for visible light utilization. • Photocatalytic activity was evaluated by degrading EDCs under visible light. • 3.0 wt% Ag/TiO 2 resulted with good photocatalytic efficiency over others. - Abstract: Surface deposition of silver nanoparticles (Ag NPs) onto the 100% anatase titania (Ag/TiO 2 ) for evolution of surface plasmon resonance (SPR) was achieved sustainably with the assistance of solar energy. The preparation resulted in Ag/TiO 2 photocatalyst with varied Ag depositions (0.5 wt%, 1.0 wt%, 3.0 wt% and 5.0 wt%). All obtained photocatalysts were characterized for the evolution of SPR via crystalline phase analysis, morphology, lattice fringes, surface area and pore size characteristics, chemical composition with chemical and electronic state, Raman scattering, optical and photoluminescence properties. The deposition of synthesized Ag NPs exhibited high uniformity and homogeneity and laid pathway for effective utilization of the visible region of electromagnetic spectrum through SPR. The depositions also lead for suppressing recombination rates of electron–hole. The photocatalytic evaluation was carried out by adopting two different class of endocrine disturbing compound (EDC) i.e., amoxicillin (pharmaceutical) and 2,4-dichlorophenol (pesticide) excited with artificial visible light source. Ag/TiO 2 with Ag > 0.5 wt% exhibited significant degradation efficiency for both amoxicillin and 2,4-dichlorophenol. Thus synthesized Ag/TiO 2 revealed the implication of plasmonics on TiO 2 for the enhanced visible light photocatalytic activity

Antibody-Hapten Recognition at the Surface of Functionalized Liposomes Studied by SPR: Steric Hindrance of Pegylated Phospholipids in Stealth Liposomes Prepared for Targeted Radionuclide Delivery

Directory of Open Access Journals (Sweden)

Eliot. P. Botosoa

2011-01-01

Full Text Available Targeted PEGylated liposomes could increase the amount of drugs or radionuclides delivered to tumor cells. They show favorable stability and pharmacokinetics, but steric hindrance of the PEG chains can block the binding of the targeting moiety. Here, specific interactions between an antihapten antibody (clone 734, specific for the DTPA-indium complex and DTPA-indium-tagged liposomes were characterized by surface plasmon resonance (SPR. Non-PEGylated liposomes fused on CM5 chips whereas PEGylated liposomes did not. By contrast, both PEGylated and non-PEGylated liposomes attached to L1 chips without fusion. SPR binding kinetics showed that, in the absence of PEG, the antibody binds the hapten at the surface of lipid bilayers with the affinity of the soluble hapten. The incorporation of PEGylated lipids hinders antibody binding to extents depending on PEGylated lipid fraction and PEG molecular weight. SPR on immobilized liposomes thus appears as a useful technique to optimize formulations of liposomes for targeted therapy.

Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia Martinez, H

1999-07-01

In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 {sup L}ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III

International Nuclear Information System (INIS)

Parra M, M. A.; Luis L, M. A.; Raya A, R.; Cruz G, H. S.

2013-10-01

This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10 -10 to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of 197 Au, 58 Ni, 115 In, 24 Mg, 27 Al, 58 Fe, 59 Co and 63 Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

Dynamic SPR monitoring of yeast nuclear protein binding to a cis-regulatory element

International Nuclear Information System (INIS)

Mao, Grace; Brody, James P.

2007-01-01

Gene expression is controlled by protein complexes binding to short specific sequences of DNA, called cis-regulatory elements. Expression of most eukaryotic genes is controlled by dozens of these elements. Comprehensive identification and monitoring of these elements is a major goal of genomics. In pursuit of this goal, we are developing a surface plasmon resonance (SPR) based assay to identify and monitor cis-regulatory elements. To test whether we could reliably monitor protein binding to a regulatory element, we immobilized a 16 bp region of Saccharomyces cerevisiae chromosome 5 onto a gold surface. This 16 bp region of DNA is known to bind several proteins and thought to control expression of the gene RNR1, which varies through the cell cycle. We synchronized yeast cell cultures, and then sampled these cultures at a regular interval. These samples were processed to purify nuclear lysate, which was then exposed to the sensor. We found that nuclear protein binds this particular element of DNA at a significantly higher rate (as compared to unsynchronized cells) during G1 phase. Other time points show levels of DNA-nuclear protein binding similar to the unsynchronized control. We also measured the apparent association complex of the binding to be 0.014 s -1 . We conclude that (1) SPR-based assays can monitor DNA-nuclear protein binding and that (2) for this particular cis-regulatory element, maximum DNA-nuclear protein binding occurs during G1 phase

RB Research nuclear reactor, Annual report for 1994, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1994. godini, I - III

Energy Technology Data Exchange (ETDEWEB)

Stefanovic, D; Milosevic, M; Pesic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia); Marinkovic, P [Elektrotehnicki fakultet, beograd (Yugoslavia); Kocic, A; Ilic, R; Dasic, N; Ljubenov, V; Petronijevic, M; Jevremovic, M [Institute of Nuclear Sciences Vinca, Belgrade (Serbia)

1994-12-15

Report on RB reactor operation during 1994 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization as well as operation of the VAX-8250 computer.

Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico

International Nuclear Information System (INIS)

Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J.

2001-01-01

This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

Extraction behaviour of Am(III) and Eu(III) from nitric acid medium in TEHDGA-HDEHP impregnated resins

Energy Technology Data Exchange (ETDEWEB)

Saipriya, G.; Kumar, T. [Bhabha Atomic Research Centre Facilities, Kalpakkam (India). Kalpakkam Reprocessing Plant; Kumaresan, R.; Nayak, P.K.; Venkatesan, K.A.; Antony, M.P. [Indira Gandhi Center for Atomic Research, Kalpakkam (India). Fuel Chemistry Div.

2016-07-01

The extraction behaviour of Am(III) and Eu(III) from nitric acid medium was studied in the solvent impregnated resins containing extractants such as tetra-bis(2-ethylhexyl)diglycolamide (TEHDGA) or bis-(2-ethylhexyl)phosphoric acid (HDEHP) or mixture of TEHDGA+HDEHP. The rate of extraction of Am(III) and Eu(III) from 1 M nitric acid and the effect of various parameters, such as the concentration of nitric acid in aqueous phase and concentration of TEHDGA and HDEHP in resin phase, on the distribution coefficient of Am(III) and Eu(III) was studied. The distribution coefficient of Am(III) and Eu(III) in HDEHP-impregnated resin decreased and that in TEHDGA-impregnated resin increased, with increase in the concentration of nitric acid. However, in (TEHDGA+HDEHP) - impregnated resin, synergic extraction was observed at lower nitric acid concentration and antagonism at higher nitric acid concentration. The mechanism of Am(III) and Eu(III) extraction in the combined resin was investigated by slope analysis method. The extraction of various metal ions present in the fast reactor simulated high-level liquid waste was studied. The separation factor of Am(III) over Eu(III) was studied using citrate-buffered diethylenetriaminepentaacetic acid (DTPA) solution.

Joining of polymer-metal lightweight structures using self-piercing riveting (SPR) technique: Numerical approach and simulation results

Science.gov (United States)

Amro, Elias; Kouadri-Henni, Afia

2018-05-01

Restrictions in pollutant emissions dictated at the European Commission level in the past few years have urged mass production car manufacturers to engage rapidly several strategies in order to reduce significantly the energy consumption of their vehicles. One of the most relevant taken action is light-weighting of body in white (BIW) structures, concretely visible with the increased introduction of polymer-based composite materials reinforced by carbon/glass fibers. However, the design and manufacturing of such "hybrid" structures is limiting the use of conventional assembly techniques like resistance spot welding (RSW) which are not transferable as they are for polymer-metal joining. This research aims at developing a joining technique that would eventually enable the assembly of a sheet molding compound (SMC) polyester thermoset-made component on a structure composed of several high strength steel grades. The state of the art of polymer-metal joining techniques highlighted the few ones potentially able to respond to the industrial challenge, which are: structural bonding, self-piercing riveting (SPR), direct laser joining and friction spot welding (FSpW). In this study, the promising SPR technique is investigated. Modelling of SPR process in the case of polymer-metal joining was performed through the building of a 2D axisymmetric FE model using the commercial code Abaqus CAE 6.10-1. Details of the numerical approach are presented with a particular attention to the composite sheet for which Mori-Tanaka's homogenization method is used in order to estimate overall mechanical properties. Large deformations induced by the riveting process are enabled with the use of a mixed finite element formulation ALE (arbitrary Lagrangian-Eulerian). FE model predictions are compared with experimental data followed by a discussion.

Informační systém pro správu rezervací v restauracích

OpenAIRE

Slanař, David

2009-01-01

Tato bakalářská práce pojednává o stávajícím stavu fungování rezervací v restauracích a jeho inovaci pomocí zavedení informačního systému pro vytváření a správu rezervací. Popisuje návrh internetové webové aplikace umožňující on-line vytváření rezervací samotnými uživateli a jejich následnou správu zaměstnanci. Dále popis ekonomického přínosu a jiných výhod plynoucích ze zavedení tohoto systému v reálném světě a v neposlední řadě řešení případných problémů při nasazování systému v praxi. T...

Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

RB research nuclear reactor, Annual report for 1989, I - III; Istrazivacki nukleani reaktor RB (Izvestaj o radu u 1989. godini), I - III

Energy Technology Data Exchange (ETDEWEB)

Stefanovic, D; Pesic, M; Hadimahmutovic, N; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1989-12-15

This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989.

Solar neutrino oscillation parameters after SNO Phase-III and SAGE Part-III

International Nuclear Information System (INIS)

Yang Ping; Liu Qiuyu

2009-01-01

We analyse the recently published results from solar neutrino experiments SNO Phase-III and SAGE Part-III and show their constraints on solar neutrino oscillation parameters, especially for the mixing angle θ 12 . Through a global analysis using all existing data from SK, SNO, Ga and Cl radiochemical experiments and long base line reactor experiment KamLAND , we obtain the parameters Δm 12 2 =7.684 -0.208 +0.212 x 10 -5 eV 2 , tan 2 θ 12 =0.440 -0.057 +0.059 . We also find that the discrepancy between the KamLAND and solar neutrino results can be reduced by choosing a small non-zero value for the mixing angle θ 13 . (authors)

RB research nuclear reactor - Annual report for 1986, I - III; Istrazivacki nuklearni reaktor RB (Izvestaj o radu u 1986. godini), I-III

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1987-07-01

This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986. Izvestaj pokazuje podatke o radu reaktora RB u toku 1986. godine, stanje reaktorske opreme, podatke o angazovanom osoblju na reaktoru i datoteku sa podacima o eksperimentalnoj i drugoj opremi reaktora RB. Sastoji se od 3 dela: tehnicki opis, pogon i rad reaktora, oprema postrojenja i njeno odrzavanje, koriscenje reaktora u 1986. godini.

Determination of the flows profile in the role of power in the central thimble of TRIGA Mark III Reactor

International Nuclear Information System (INIS)

Garcia F, A.

2010-01-01

The overall objective of the thesis project is to determine the flow profiles sub cadmic and epi cadmic in the central thimble to different powers and operation times of TRIGA Mark III Reactor, using activation foils as detectors. In the reactor operation, it is necessary to know the neutron flow profile for to realize other tasks as: the radioisotopes production, research in reactors physics and fuel burning. The distribution of the neutron flow, accurately reflects what is happening in the reactor core, plus the flows value in this distribution is directly related to the power generated. For this reason it is performed the sub cadmic flow measurement with energies between 0 and 0.4 eV (energy of the cadmium cut E cd ∼ 0.4 eV) and epi cadmic flow with energies greater than 0.4 eV, in the central thimble powers to the powers of 10, 100 W, 1, 10 100 Kw and 1 MW. The method used is known as flakes activation, which is to be arranged by placing flakes ( 3 mm of diameter and 0.0508 mm of thickness) of a given material (either Au, In, Cu, Mn, etc.) into an aluminum tube outside diameter equal to 6.35 mm, alternating flakes with lids covered and discovered of cadmium (3.4 mm of diameter and 0.508 mm of thickness) and separated by lucite pieces of 3 mm of diameter and 25.4 mm in length. After irradiating the flakes for some time, is measured the gamma activity of each of them, using a hyper pure germanium detector of high resolution. Already known gamma activity, proceed to calculate the epi cadmic and sub cadmic flows using a computer program in Fortran language, called Caflu. (Author)

Behavior of exposed human lymphocytes to a neutron beam of the Reactor TRIGA Mark III; Comportamiento de linfocitos humanos expuestos a un haz de neutrones del Reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Carbajal R, M. I.; Arceo M, C.; Aguilar H, F.; Guerrero C, C., E-mail: citlali.guerrero@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2012-10-15

The living beings are permanently exposed to radiations of natural origin: cosmic and geologic, as well as the artificial radiations that come from sources elaborated by the man. The artificial sources have an important use in the medical area. Particularly has been increased the neutrons use due to the effectiveness that they have to damage the cells with regard to other radiation types. The biological indicator of exposition to ionizing radiation more reliable is the chromosomal aberrations study, specifically the dicentrics in human lymphocytes. This test allows, establishing the exposition dose in function of the damage quantity. The dicentrics have a behavior in function of the dose. The calibration curve that describes this behavior is specific for each type of ionizing radiation. In the year 2006 beginning was given to the expositions of human lymphocytes to a neutron beam generated in the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico. Up to 2008 the response dose curve comprised an interval of exposition time of up to 30 minutes. Moreover, the interval between 10 an 20 minutes is included, since was observed that this last is indispensable for the adjustment waited in a lineal model. (Author)

Burnup calculation for a tokamak commercial hybrid reactor

International Nuclear Information System (INIS)

Feng Kaiming; Xie Zhongyou

1990-08-01

A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ

International Nuclear Information System (INIS)

Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F.

2007-01-01

The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

Research reactor`s role in Korea

Energy Technology Data Exchange (ETDEWEB)

Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1996-12-31

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

Development of a software for the control of the quality management system of the TRIGA-Mark III reactor; Desarrollo de un software para el control del sistema de gestion de calidad del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Herrera A, E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Hernandez, L.V.; Hernandez, J.A. [UAEM, Depto. de Ingenieria en Computacion, 50000 Toluca, estado de Mexico (Mexico)]. e-mail: eha@nuclear.inin.mx

2006-07-01

The quality has not only become one of the essential requirements of the product but rather at the presenme it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

SIMMER-III code-verification. Phase 1

International Nuclear Information System (INIS)

Maschek, W.

1996-05-01

SIMMER-III is a computer code to investigate core disruptive accidents in liquid metal fast reactors but should also be used to investigate safety related problems in other types of advanced reactors. The code is developed by PNC with cooperation of the European partners FZK, CEA and AEA-T. SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-, time-, and energy-dependent neutron dynamics model. In order to model complex flow situations in a postulated disrupting core, mass and energy conservation equations are solved for 27 density components and 16 energy components, respectively. Three velocity fields (two liquid and one vapor) are modeled to simulate the relative motion of different fluid components. An additional static field takes into account the structures available in a reactor (pins, hexans, vessel structures, internal structures etc.). The neutronics is based on the discrete ordinate method (S N method) coupled into a quasistatic dynamic model. The code assessment and verification of the fluid dynamic/thermohydraulic parts of the code is performed in several steps in a joint effort of all partners. The results of the FZK contributions to the first assessment and verification phase is reported. (orig.) [de

Comparative study of SPR and ELISA methods based on analysis of CD166/ALCAM levels in cancer and control human sera

Czech Academy of Sciences Publication Activity Database

Vaisocherová, Hana; Faca, V.M.; Taylor, A. D.; Hanash, S.; Jiang, S.

2009-01-01

Roč. 24, č. 7 (2009), s. 2143-2148 ISSN 0956-5663 Institutional research plan: CEZ:AV0Z20670512 Keywords : SPR * ELISA * Cancer biomarkers Subject RIV: CE - Biochemistry Impact factor: 5.429, year: 2009

Plasma processing of the Si(0 0 1) surface for tuning SPR of Au/Si-based plasmonic nanostructures

Energy Technology Data Exchange (ETDEWEB)

Giangregorio, Maria M. [Institute of Inorganic Methodologies and of Plasmas, IMIP-CNR and INSTM sez. Bari, Via Orabona 4, 70125 Bari (Italy)]. E-mail: michelaria.giangregorio@ba.imip.cnr.it; Losurdo, Maria [Institute of Inorganic Methodologies and of Plasmas, IMIP-CNR and INSTM sez. Bari, Via Orabona 4, 70125 Bari (Italy); Sacchetti, Alberto [Institute of Inorganic Methodologies and of Plasmas, IMIP-CNR and INSTM sez. Bari, Via Orabona 4, 70125 Bari (Italy); Capezzuto, Pio [Institute of Inorganic Methodologies and of Plasmas, IMIP-CNR and INSTM sez. Bari, Via Orabona 4, 70125 Bari (Italy); Bruno, Giovanni [Institute of Inorganic Methodologies and of Plasmas, IMIP-CNR and INSTM sez. Bari, Via Orabona 4, 70125 Bari (Italy)

2006-12-15

Au nanoclusters have been deposited on Si(0 0 1) surfaces by sputtering of a metallic Au target using an Ar plasma. Different wet and dry treatments of the Si(0 0 1) surface, including dipping in HF solution and exposure to H{sub 2} and N{sub 2} plasmas, have been applied and the effects of these treatments on the Au nanoparticles/Si interface, the Au nanoclusters aspect ratio and the surface plasmon resonance (SPR) energy and amplitude are investigated exploiting spectroscopic ellipsometry and atomic force microscopy. It is found that the Au nanoclusters aspect ratio depends on the extent of the Au-Si intermixing. The thicker the Au-Si interface layer, the larger the Au nanoparticles aspect ratio and the red-shift of the SPR peak. Furthermore, SiO{sub 2} and the H{sub 2} plasma treatment inhibit the Si-Au intermixing, while HF-dipping and the N{sub 2} plasma treatment favour Au-Si intermixing, yielding silicide formation which increases the Si wetting by Au.

Thermo-mechanical design and structural analysis of the first wall for ARIES-III, A 1000 MWeD-3He power reactor

International Nuclear Information System (INIS)

Sviatoslavsky, I.; Blanchard, J.P.; Mogahed, E.A.

1992-01-01

This paper reports on ARIES III, a conceptual design study of a 1000 MWe D- 3 He tokamak fusion power reactor in which most of the energy comes from charged particle transport, bremsstrahlung and synchrotron radiation, and only a small fraction (∼ 4%) comes form neutrons. This form of energy is deposited as surface heating on the chamber first wall (FW) and divertor elements, while the neutron energy is deposited as bulk nuclear heating within the shield. Since this reactor does not use tritium, there is no breeding blanket. Instead a shield is provided to protect the magnets from neutrons. The Fw is very unique in a D- 3 He reactor, it must be capable of absorbing the high surface heat in a mode suitable for efficient power cycle conversion, it must be able to reflect synchrotron radiation, and it must be able to withstand high current plasma disruptions. The FW is made of a low activation ferritic steel (MHT-9) and is cooled with an organic coolant (HB-40) at a pressure of 2 MPa. The FW has a coating of 0.01 cm tungsten on the MHT-9, followed by 0.15 cm of Be on the plasma side. This is needed for synchrotron radiation reflection and as a melt layer to guard against the thermal effects of a plasma disruption

SGN III code conversion from Apollo to HP

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-04-01

SGN III computer code is used to analyze transient behavior of reactor coolant system, pressurizer and steam generators in the event of main steam line break (MSLB), and to calculate mass and energy release for containment design. This report firstly describes detailed work carried out for installation of SFN III on Apollo DN 10000 and code validation results after installation. Secondly, a series of work is also describes in relation to installation of SGN III on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 8 refs. (Author) .new

Firebird-III program description

International Nuclear Information System (INIS)

Lin, M.R.; Prawirosochardjo, S.; Rennick, D.F.; Wessman, E.; Blain, R.J.D.; Wilson, J.M.

1979-09-01

The FIREBIRD-III digital computer program is a general network code developed primarily for predicting the thermalhydraulic behaviour of CANDU power reactors during a postulated loss-of-coolant accident and the subsequent emergency coolant injection. Because of its flexibility, the code can also be used to solve a large variety of general two-phase flow problems. This report describes the thermalhydraulic models and the computation methods used in the program

Tritium management in fusion reactors

International Nuclear Information System (INIS)

Galloway, T.R.

1978-05-01

This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

Energy Technology Data Exchange (ETDEWEB)

Barranco R, F.

2015-07-01

In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

Single-phase and two-phase anaerobic digestion of fruit and vegetable waste: Comparison of start-up, reactor stability and process performance

Energy Technology Data Exchange (ETDEWEB)

Ganesh, Rangaraj [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Torrijos, Michel, E-mail: michel.torrijos@supagro.inra.fr [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Sousbie, Philippe [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Lugardon, Aurelien [Naskeo Environnment, 52 rue Paul Vaillant Couturier, F-92240 Malakoff (France); Steyer, Jean Philippe; Delgenes, Jean Philippe [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France)

2014-05-01

Highlights: • Single-phase and two-phase systems were compared for fruit and vegetable waste digestion. • Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS and 83% VS removal. • Substrate solubilization was high in acidification conditions at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2. • Energy yield was lower by 33% for two-phase system compared to the single-phase system. • Simple and straight-forward operation favored single phase process over two-phase process. - Abstract: Single-phase and two-phase digestion of fruit and vegetable waste were studied to compare reactor start-up, reactor stability and performance (methane yield, volatile solids reduction and energy yield). The single-phase reactor (SPR) was a conventional reactor operated at a low loading rate (maximum of 3.5 kg VS/m{sup 3} d), while the two-phase system consisted of an acidification reactor (TPAR) and a methanogenic reactor (TPMR). The TPAR was inoculated with methanogenic sludge similar to the SPR, but was operated with step-wise increase in the loading rate and with total recirculation of reactor solids to convert it into acidification sludge. Before each feeding, part of the sludge from TPAR was centrifuged, the centrifuge liquid (solubilized products) was fed to the TPMR and centrifuged solids were recycled back to the reactor. Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS fed and VS removal of 83%. The TPAR shifted to acidification mode at an OLR of 10.0 kg VS/m{sup 3} d and then achieved stable performance at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2, with very high substrate solubilization rate and a methane yield of 0.30 m{sup 3} CH{sub 4}/kg COD fed. The two-phase process was capable of high VS reduction, but material and energy balance showed that the single-phase process was superior in terms of volumetric methane production and energy yield by 33%. The lower energy yield of the two-phase system was due to the loss of

Single-phase and two-phase anaerobic digestion of fruit and vegetable waste: Comparison of start-up, reactor stability and process performance

International Nuclear Information System (INIS)

Ganesh, Rangaraj; Torrijos, Michel; Sousbie, Philippe; Lugardon, Aurelien; Steyer, Jean Philippe; Delgenes, Jean Philippe

2014-01-01

Highlights: • Single-phase and two-phase systems were compared for fruit and vegetable waste digestion. • Single-phase digestion produced a methane yield of 0.45 m 3 CH 4 /kg VS and 83% VS removal. • Substrate solubilization was high in acidification conditions at 7.0 kg VS/m 3 d and pH 5.5–6.2. • Energy yield was lower by 33% for two-phase system compared to the single-phase system. • Simple and straight-forward operation favored single phase process over two-phase process. - Abstract: Single-phase and two-phase digestion of fruit and vegetable waste were studied to compare reactor start-up, reactor stability and performance (methane yield, volatile solids reduction and energy yield). The single-phase reactor (SPR) was a conventional reactor operated at a low loading rate (maximum of 3.5 kg VS/m 3 d), while the two-phase system consisted of an acidification reactor (TPAR) and a methanogenic reactor (TPMR). The TPAR was inoculated with methanogenic sludge similar to the SPR, but was operated with step-wise increase in the loading rate and with total recirculation of reactor solids to convert it into acidification sludge. Before each feeding, part of the sludge from TPAR was centrifuged, the centrifuge liquid (solubilized products) was fed to the TPMR and centrifuged solids were recycled back to the reactor. Single-phase digestion produced a methane yield of 0.45 m 3 CH 4 /kg VS fed and VS removal of 83%. The TPAR shifted to acidification mode at an OLR of 10.0 kg VS/m 3 d and then achieved stable performance at 7.0 kg VS/m 3 d and pH 5.5–6.2, with very high substrate solubilization rate and a methane yield of 0.30 m 3 CH 4 /kg COD fed. The two-phase process was capable of high VS reduction, but material and energy balance showed that the single-phase process was superior in terms of volumetric methane production and energy yield by 33%. The lower energy yield of the two-phase system was due to the loss of energy during hydrolysis in the TPAR and the

Sambandet mellan Theory of Mind, språkliga förmågor och exekutiva funktioner hos barn i treårsåldern

OpenAIRE

Nilsson, Anna; Sirén, Ellen

2015-01-01

Barns fem första levnadsår präglas av en omfattande utveckling av språkliga förmågor och exekutiva funktioner, vilka är färdigheter som visat sig ha en stor påverkan på hur barn utvecklar förståelse för sina egna och andras tankar, det vill säga deras Theory of Mind. En central aspekt inom forskning har varit att undersöka hur förmågorna samvarierar vid olika åldrar. Syftet i föreliggande studie var att undersöka sambandet mellan Theory of Mind, språkliga förmågor och exekutiva funktioner hos...

Pre-Preliminary results from the phase III of the IAEA CRP: optimizing of reactor pressure vessel surveillance programmes and their analysis

Energy Technology Data Exchange (ETDEWEB)

Brumovsky, M; Gillemot, F; Kryukov, A; Levit, V

1994-12-31

This paper gives preliminary results and some conclusions from Phase III of the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses`` carried out during the last seven years in 15 member states. First analysis concerned: comparison of results from initial, un-irradiated materials condition, comparison of transition temperature shifts (from notch toughness testing) with respect to content of residual (P, Cu) and alloying (Ni) elements, type of material (base and weld metal), irradiation temperature (288 and 265 C), and type of fluence dependence. Special effort has been taken to the analysis of the behaviour of a chosen reference steel. (JRQ). 6 figs., 4 tabs.

Preliminary studies for the upgrading of the ININ TRIGA MARK III reactor

International Nuclear Information System (INIS)

Aguilar, J.V.

1984-01-01

An increase on the demand of services has led to the -consideration of reactor upgrading, for which a project has been proposed. This project includes the modification of the reactor core, using FLIP type (Erbium loaded, 70% U-235 enriched) fuel elements and the upgrading of the control - console which implies a careful analysis of the Safety -considerations involved in the system modification. In this paper, a brief description is given of: The modifications intended; Pre-operational tests for the system components to be modified; The safety report needed for the licensing of the upgraded reactor; The project management and evaluation

Theoretical evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core

International Nuclear Information System (INIS)

Paredes G, L.C.

1991-11-01

It was theoretically determined the accumulation of the Xe 135 and Sm 149 in function, of the time during a stationary state of 72 h. continuous for the reactor TRIGA Mark III to 1 MW of thermal power with mixed core. The values of negative reactivity due to these isotopes are of 2.04 dollars and 0.694 dollars to the 72 h, quantities that will have to be compensated if wants that the reactor continues working to this power. Under the same conditions but considering a core with standard fuel, it was found a value of ρ = 1.70 dollars, resulting a difference of 0.30 dollars of negative reactivity in function of the type of analyzed core. This difference is important for the calculations of fuel management of a reactor. The concentration in balance of the xenon was reaches after an operation to constant power of 1 MW by 50 h, contrary to the samarium that reaches it balance after 3 weeks of operation starting from the initial start up and it stays constant along the useful life of the reactor while a change of fuel doesn't exist. It was obtained that for operation times greater to 60 h. at 1 MW, a peak of negative reactivity of the Xe 135 is generated between the 7 and 11 h after the instantaneous shut down, with a value of 2.43 dollars, that is to say 0.39 additional dollars to those taken place during the continuous irradiation. (Author)

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)

Recent progress in safety-related applications of reactor noise analysis

International Nuclear Information System (INIS)

Hirota, Jitsuya; Shinohara, Yoshikuni; Saito, Keiichi

1982-01-01

Recent progress in safety-related applications of reactor noise analysis is reviewed, mainly referring to various papers presented at the Third Specialists' Meeting on Reactor Noise (SMORN-III) held in Tokyo in 1981. Advances in application of autoregressive model, coherence analysis and pattern recognition technique are significant since SMORN-II in 1977. Development of reactor diagnosis systems based on noise analysis is in progress. Practical experiences in the safety-related applications to power plants are being accumulated. Advances in quantitative monitoring of vibration of internal structures in PWR and diagnosis of core stability and control system characteristics in BWR are notable. Acoustic methods are also improved to detect sodium boiling in LMFBR. The Reactor Noise Analysis Benchmark Test performed by Japan in connection with SMORN-III is successful so that it is possible to proceed to the second stage of the benchmark test. (author)

An overview of future sustainable nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

2013-07-01

In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will tend to have closed

Upgrade of reactor operation technology

International Nuclear Information System (INIS)

Itoh, Hideaki; Suzuki, Toshiaki; O-kawa, Toshikatsu

2003-01-01

To improve operational reliability and availability, the operation technology for a fast reactor was developed in the ''JOYO''. This report describes the upgrading of the simulator, plant operation management tools and fuel handling system for the MK-III core operation. The simulator was modified to the MK-III version to verify operation manuals, and to train operators in MK-III operation. The plant operation management tool was replaced on the operation experience to increase the reliability and efficiency of plant management works relating to plant operation and maintenance. To shorten the refueling period, the fuel handling system was upgraded to full automatic remote control. (author)

Theoretical analysis of nuclear reactors (Phase II), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, III Deo, Zatrovanje reaktora, II faza

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-10-15

This phase is dealing with influence of all the fission products except Xe{sup 135} on the reactivity of a reactor, usually named as reactor poisoning. The first part of the report is a review of methods for calculation of reactor poisoning. The second part shows the most frequently used method for calculation of cross sections and yields of pseudo products (for thermal neutrons). The system of equations was adopted dependent on the conditions of the available computer system. It is described in part three. Detailed method for their application is described in part four and results obtained are presented in part five.

Reforma veřejné správy: analýza Strategie realizace Smart Administration

OpenAIRE

Hurychová, Tereza

2016-01-01

This diploma thesis is mainly to explain the significance of the document Efektivní veřejná správa a přátelské služby: Strategie realizace Smart Administration 2007 - 2015 2015 (Efficient public administration and friendly service: Smart Administration Strategy 2007 - 2015) in the context of the reform process of public administration in the Czech republic. The author′s point of view is based of the theory of change, strategic managment and modern managment methods in public administration. A...

Summary of the progress of reactor physics in Japan reviewing the activities related to NEA Committee on Reactor Physics

International Nuclear Information System (INIS)

Hirota, Jitsuya

1984-09-01

The progress of fast and thermal reactor physics, fusion neutronics and shielding researches in these twenty years can be clearly recognized in the reviews of reactor physics activities in Japan which had been perpared by the Special Committee on Reactor Physics: the joint committee under Atomic Energy Society of Japan and JAERI. Many topics of those discussed at the NEACRP meetings concerned fast reactor physics. Information exchange on the topics such as adjustment of group cross sections by integral data, central worth discrepancy, sodium void effect and heterogeneous core stimulated the researches in Japan. And achievements in Japan including those in the JAERI Fast Critical Facility FCA were reported and contributed largely to the international co-operation. In addition, the contribution from Japan was also made concerning a study of fusion blanket. Among various specialists' meetings recommended by NEACRP, those on nuclear data and benchmarks for reactor shielding were often held since 1973 and helpful to the progress of shielding researches in Japan. The Third Specialists' Meeting on Reactor Noise (SMORN-III) was held in Tokyo in 1981, indicating the recent progress in safety-related applications of reactor noise analysis. The NEACRP benchmark tests were quite useful to the progress of reactor physics in Japan, which included the benchmark calculations of BWR lattice cell, key parameters and burn-up characteristics of a large LMFBR, FBR and PWR shielding, and so on. It may be noted that the benchmark test on reactor noise analysis methods was successfully conducted by Japan in connection with SMORN-III. In addition, the co-operation was positively made to the compilation of light water lattice data, and the preparation of reviews on actinide production and burn-up, and blanket physics. (J.P.N.)

Real-time monitoring of mycobacterium genomic DNA with target-primed rolling circle amplification by a Au nanoparticle-embedded SPR biosensor.

Science.gov (United States)

Xiang, Yang; Zhu, Xiaoyan; Huang, Qing; Zheng, Junsong; Fu, Weiling

2015-04-15

In this study, we developed a surface plasmon resonance (SPR) DNA biosensor array based on target-primed rolling circle amplification (RCA) for isothermal and rapid detection of two pathogenic mycobacteria, Mycobacterium tuberculosis complex (MTBC) and Mycobacterium avium complex (MAC).The species-specific padlock probe (PLP) was designed to target the sequence in 16S-23S rRNA gene internal transcribed spacer (ITS). After ligation, the circularized PLP could be primed by the target sequence to initial RCA. The RCA performed simultaneously with the cleavage reaction to produce small fragments of single strand DNA which immediately hybridized with the probe immobilized on the sensor chip without denaturation. This process caused SPR angle changes on the chip surface, which made the detection for analysis from the solution achievable, and dynamic real-time RCA monitoring of mycobacterium possible. Besides, Au nanoparticles (AuNPs) were directly assembled onto the surface of the sensor chip via hexanedithiol (HDT) for the enhancement of sensitivity as a label-free detection system. Experimental results show that the signal enhancement by the target-primed RCA together with AuNPs-embedded surface caused at least10-fold increased sensitivity as compared with conventional RCA on bare SPR chip method. Within 40min amplification duration as low as 20amol of synthetic targets and 10(4)CFUmL(-1) of genomic DNA from clinical samples can be detected. The proposed method not only provides a simple design idea for liquid-phase amplification monitoring, but also apply it in clinical pathogen detection, which holds great promise in ultrasensitive bioassay in the future. Copyright © 2014. Published by Elsevier B.V.

Proceedings (slides) of the OECD/NEA Workshop on Innovations in Water-cooled Reactor Technologies

International Nuclear Information System (INIS)

Spiler, Joze; Kim, Sang-Baik; ); Feron, Fabien; Jaervinen, Marja-Leena; Husse, Julien; ); Ferraro, Giovanni; Bertels, Frank; Denk, Wolfgang; Tuomisto, Harri; Golay, Michael; Buongiorno, J.; Todreas, N.; Adams, E.; Briccetti, A.; Jurewicz, J.; Kindfuller, V.; Srinivasan, G.; Strother, M.; Minelli, P.; Fasil, E.; Zhang, J.; Genzman, G.; Epinois, Bertrand de l'; Kim, Shin Whan; Laaksonen, Jukka; Maltsev, Mikhail; Yu, CHongxing; Powell, David; Gorgemans, Julie; Hopwood, Jerry; Bylov, Igor; Bakhmetyev, Alexander M.; Lepekhin, Andrey N.; Fadeev, Yuriy P.; Bruna, Giovanni; Gulliford, Jim; ); Ham-Su, Rosaura; Thevenot, Caroline; GAUTIER, Guy-Marie; MARSAULT, Philippe; PIGNATEL, Jean-Francois; White, Andrew; )

2015-02-01

New technologies and solutions have been developed over more than thirty years to improve the safety, performance and economics of nuclear power plants. Particular efforts were made in designing systems to prevent or mitigate nuclear accidents and, greatly limit or even avoid any offsite release of radioactivity. Reactor designs developed in the 1980's and later are often referred to as Generation III (Gen III) reactors. They offer enhanced safety compared to earlier Generation II (Gen II) designs, as well as improved performance and economics. Examples of Gen III safety design features include solutions for corium localisation, advanced containment structures, improved emergency core-cooling systems, filtered venting systems, hydrogen risk management solutions, etc. Some of these solutions have also been back-fitted or partially adapted to existing reactors, based on recommendations from regulators or modernisation efforts by the utilities operating these reactors, to bring their level of safety to levels approaching those of the more modern designs. Other innovations found in the latest water-cooled reactor designs include the use of passive safety systems, and often associated with those, a simplification in the design of the reactor. Gen III reactors also feature better economics, for example increased design lifetime up to 60 years, ability to use 100% MOX fuel and operate with higher flexibility, higher thermal efficiencies and reduced staff requirements. Modularity is often quoted as a feature of some Gen III designs as a way of reducing the construction times and simplifying the decommissioning of the plant. The scope of the Workshop includes, inter alia: - Evolution of regulatory and design requirements for commercial water-cooled reactors; - Innovations in water-cooled reactor technologies that allowed significant improvement in the level of safety, with a discussion on advantages and challenges of active vs. passive safety systems; - Innovations under

Decommissioning of the Neuherberg Research Reactor (FRN)

International Nuclear Information System (INIS)

Demmeler, M.; Rau, G.; Strube, D.

1982-01-01

The Neuherberg Research Reactor is of type TRIGA MARK III with 1 MW steady state power and pulsable up to 2000 MW. During more than ten years of operation 12000 MWh and 6000 reactor pulses had been performed. In spite of its good technical condition and of permanent safe operation without any failures, the decommissioning of the Neuherberg research reactor was decided by the GSF board of directors to save costs for maintaining and personnel. As the mode of decommissioning the safe enclosure was chosen which means that the fuel elements will be transferred back to the USA. All other radioactive reactor components will be enclosed in the reactor block. Procedures for licensing of the decommissioning, dismantling procedures and time tables are presented

RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

International Nuclear Information System (INIS)

2015-01-01

The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor; Diseno y construccion del SIPPING para combustibles del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

2003-07-01

The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10

International Nuclear Information System (INIS)

1984-02-01

This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff-approved acceptance criteria for LOCA-related hydrodynamic loads are provided in Appendix C of this report

Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

International Nuclear Information System (INIS)

Cai Jianping; Shen Sen; Barkman, N.

1999-01-01

The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied

Support to the elaboration of the engineering of detail, configuration and programming of the control system of heat removal of the TRIGA Mark III reactor; Apoyo a la elaboracion de la ingenieria de detalle, configuracion y programacion del sistema de control de remocion de calor del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Diaz G, C. A.

2016-07-01

Nowadays, the peaceful and responsible use of nuclear energy in Mexico is of great importance and contributes to economic, social, scientist and technologic development in the country, highlighting the Instituto Nacional de Investigaciones Nucleares (ININ) and the Nuclear Power Plant of Laguna Verde as one of the most important dependences. Among the main facilities and laboratories of ININ is the Nuclear Research Reactor TRIGA Mark III, this is a pool type reactor with mobile core, cooled and moderated by light water and a flow of 1013 n/cm{sup 2}/sec. Due to the technological obsolescence is a growing problem that threatens the information, operation and/or efficacy of elements of control and safety systems of the reactor, these must be changed each time more frequently. In the modernization of reactor was used a Modicon M340 programmable logic control (PLC) and a Twido PLC for the control of heat removal system (Primary Cooling System (PCS) and Secondary Cooling System (SCS) respectively), this because the PLC has proven to be safe and effective devices, addition to reduce the wiring elements and increase the possibilities of performance and design of the digital control console. This document shows and describes the elements of heat removal system (PCS and SCS), and the signals and signal types that such items send or received by the PLC, likewise, is indicated the methodology used to develop the applications for the control of the Primary Cooling System and Secondary Cooling System, beginning with the PLC design, the development of PLC plans and the control logic, and finally, the simulation and debugging of applications on Unity Pro and Twido Suite. All this in compliance with the safety standards to nuclear research reactors (NS-R-4), the rules of industrial programming (IEC 61131-3), and the reactor operating limits postulated in the safety report and the software assurance system used in the ININ. (Author)

Three dimensional simulation for bayou choctaw strategic petroleum reserve (SPR).

Energy Technology Data Exchange (ETDEWEB)

Ehgartner, Brian L. (Sandia National Laboratories, Albuquerque, NM); Park, Byoung Yoon; Lee, Moo Yul

2006-12-01

Three dimensional finite element analyses were performed to evaluate the structural integrity of the caverns located at the Bayou Choctaw (BC) site which is considered a candidate for expansion. Fifteen active and nine abandoned caverns exist at BC, with a total cavern volume of some 164 MMB. A 3D model allowing control of each cavern individually was constructed because the location and depth of caverns and the date of excavation are irregular. The total cavern volume has practical interest, as this void space affects total creep closure in the BC salt mass. Operations including both cavern workover, where wellhead pressures are temporarily reduced to atmospheric, and cavern enlargement due to leaching during oil drawdowns that use water to displace the oil from the caverns, were modeled to account for as many as the five future oil drawdowns in the six SPR caverns. The impacts on cavern stability, underground creep closure, surface subsidence, infrastructure, and well integrity were quantified.

SPR Characteristics Curve and Distribution of Residual Stress in Self-Piercing Riveted Joints of Steel Sheets

OpenAIRE

Haque, Rezwanul; Wong, Yat C.; Paradowska, Anna; Blacket, Stuart; Durandet, Yvonne

2017-01-01

Neutron diffraction was used to describe the residual stress distributions in self-piercing riveted (SPR) joints. The sheet material displayed a compressive residual stress near the joint, and the stress gradually became tensile in the sheet material far away from the joint. The stress in the rivet leg was lower in the thick joint of the softer steel sheet than in the thin joint of the harder steel sheet. This lower magnitude was attributed to the lower force gradient during the rivet flaring...

Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1962-07-01

Method was developed for calculation of Xe 135 static effect and kinetic effects of Xe 135 and Sm 149 with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe 13 5 microscopic capture cross section dependent on temperature [sr

Effluent releases at the TRIGA reactor facility

Energy Technology Data Exchange (ETDEWEB)

Whittemore, W L [General Atomic Co., San Diego, CA (United States)

1974-07-01

The principal effluent from the operating TRIGA reactors in our facility is argon-41. As monitored by a recording gas and particulate stack monitor, the values shown in the table, the Mark III operating 24 hours per day for very long periods produced the largest amount of radioactive argon. The quantity of 23.7 Ci A-41 when diluted by the normal reactor room ventilation system corresponded to 1.45 x 10{sup -6} {mu}Ci/cc. As diluted in the roof stack stream and the reactor building wake, the concentration immediately outside the reactor building was 25% MPC for an unrestricted area. The continued dilution of this effluent resulted in a concentration of a few percent MPC at the site boundary (unrestricted area) 350 meters from the reactor. (author)

Determination of the fluence profile in three dimension for the thermal column of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Herrera A, E.; Urena N, F.; Delfin L, A.; Garcia M, T.

2006-01-01

In this work the results of the dosimetric properties of the lithium carbonate are presented (detecting), before the thermal neutrons. The process consists on irradiating samples of lithium carbonate in the installation of the thermal column of the TRIGA Mark III reactor, with a controlled period and with time intervals of 20 hours of irradiation. It is necessary to mention that the detectors were placed in different internal positions of the thermal column. With the purpose of being used these results for future studies, like it is the fluence profile in the thermal column. To use the BNCT technique (Boron Neutron Capture Therapy). Which is a binary technique that requires the simultaneous presence of a neutron flux with appropriate energy and a neutron captor (10B), those which interacting to attack to the tumor cells without producing significant damage to the tissues when both agents are separated. (Author)

Operation and maintenance of the RA reactor in 1964, I-II, Part II; Pogon i odrzavanje reaktora RA u 1964. godini, I-II, II Deo

Energy Technology Data Exchange (ETDEWEB)

Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1964-12-15

This volume of the report contains the following 15 Annexes: Improvement of the fuel cycle economy (record No. 37009803 in INIS DB); Analysis of neutron flux increase in horizontal experimental channels of the RA reactor record No. 37005698 in INIS DB); Application of the critical system for determining the thermal neutron flux in a research reactor with central horizontal reflector ( record No. 37055005 in INIS DB); Determining the capacity of the RA reactor heat exchanger dependent on the coolant water temperature and flow; Operation of the RA reactor in forced regime; Analysis of the CEN-132 heavy water pumps failures at the RA reactor from decontamination till present; Modifications in the vacuum loop of the distillation system; Report on decontamination of the evaporator and cleaning of the condenser of the distillation system; Operation of reactor at nominal power with reduced D{sub 2}O circulation; Cooling of the RA reactor with reduced flow rate in the heavy water loop; Measurement of the heavy water level in the fuel channels of the RA reactor; Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964; Conclusions of the experts group at the meeting held on November 23 1964; After heat and the cooling problem after RA reactor shut-down; Measurement of noise and vibrations on the Ra reactor heavy water system; Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor; Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor; Study of the modifications in the synchronous generators, heavy water pumps and condenser batteries of the RA reactor.

Analysis of cavern stability at the West Hackberry SPR site.

Energy Technology Data Exchange (ETDEWEB)

Ehgartner, Brian L.; Sobolik, Steven Ronald

2009-05-01

This report presents computational analyses that simulate the structural response of caverns at the Strategic Petroleum Reserve (SPR) West Hackberry site. The cavern field comprises 22 caverns. Five caverns (6, 7, 8, 9, 11) were acquired from industry and have unusual shapes and a history dating back to 1946. The other 17 caverns (101-117) were leached according to SPR standards in the mid-1980s and have tall cylindrical shapes. The history of the caverns and their shapes are simulated in a three-dimensional geomechanics model of the site that predicts deformations, strains, and stresses. Future leaching scenarios corresponding to oil drawdowns using fresh water are also simulated by increasing the volume of the caverns. Cavern pressures are varied in the model to capture operational practices in the field. The results of the finite element model are interpreted to provide information on the current and future status of subsidence, well integrity, and cavern stability. The most significant results in this report are relevant to Cavern 6. The cavern is shaped like a bowl with a large ceiling span and is in close proximity to Cavern 9. The analyses predict tensile stresses at the edge of the ceiling during repressuization of Cavern 6 following workover conditions. During a workover the cavern is at low pressure to service a well. The wellhead pressures are atmospheric. When the workover is complete, the cavern is repressurized. The resulting elastic stresses are sufficient to cause tension around the edge of the large ceiling span. With time, these stresses relax to a compressive state because of salt creep. However, the potential for salt fracture and propagation exists, particularly towards Cavern 9. With only 200 ft of salt between the caverns, the operational consequences must be examined if the two caverns become connected. A critical time may be during a workover of Cavern 9 in part because of the operational vulnerabilities, but also because dilatant damage is

Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident; Analiza sigurnosti rada reaktora RA - I-III, III deo - Posledica maksimalno moguceg akcidenta na okolinu reaktora

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects.

Experimental evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core, configuration No. 16 (Final report of the project); Evaluacion experimental de la produccion de los venenos Xe-135 y Sm-149 del reactor TRIGA Mark III con nucleo mixto, config. No. 16 (Informe final del proyecto)

Energy Technology Data Exchange (ETDEWEB)

Paredes G, L C

1991-11-15

It was generated the concentration curve of the Xe{sup 135} (t) during the TRIGA Mark III reactor operation cycle, for a continuous irradiation of 72 h to 1 MW of thermal power, as well as the accumulation curve of the isotope after the shutdown, for the fuel configuration No. 16 in the thermal column. The maximum negative reactivities generated by the Xe{sup 135} for operation times greater than 60 h to 1 MW and after the reactor shutdown its were of 1.968 {+-} 0.15 dollars and 2.30 {+-} 0.15 dollars respectively. When comparing these results with those theoretically calculated we find differences of the order of 3.6% and 5.34% which are understood inside the experimental error that on the average was of 7.6%. The results before mentioned have an important application during the start up process of the Reactor, when analyzing the value of the weekly reactivity excess of the core and when is choice the pattern of bars to use for experiments of but of 2 h, where is required to minimize the temporary and space interferences of the neutron flux. (Author)

Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

Operation and maintenance of the RA reactor in 1964, I-II, Part I

International Nuclear Information System (INIS)

Pavicevic, M.

1964-12-01

During 1964, the Reactor as operated about 20 days each months at nominal power of 6.5 MW, 5 days at lower power levels and 5 days were used for maintenance. Total production was 27930 MWh which is 11.7% higher than the planned value. Fuel exchange was done 3 times during this period, 98 spent fuel channels were exchanged. In addition to routine maintenance of reactor components and instruments a series of analyses of heavy water and helium were done. Special attention was devoted to corrosion analyses of the reactor materials because of the heavy water system was refurbished decontaminated in 1963. Utilization of the experimental space in the reactor was better that previously. 546 samples were irradiated till the end of November, of which 443 for users from the Institute. Specific irradiations in the fast neutron flux were done in six VISA-2 channels in the core

Louisiana SIP: LAC 33:III Ch 2147. Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 2011-08-04 (LAd34) to 2017-09-27

Science.gov (United States)

Louisiana SIP: LAC 33:III Ch 2147. Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 2011-08-04 (LAd34) to 2017-09-27

Preliminary nuclear power reactor technology qualitative assessment for Malaysia

International Nuclear Information System (INIS)

Shamsul Amri Sulaiman

2011-01-01

Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

Preparation and characterization of thin organosilicon films deposited on SPR chip

Energy Technology Data Exchange (ETDEWEB)

Szunerits, Sabine [Laboratoire d' Electrochimie et de Physicochimie des Materiaux et des Interfaces (LEPMI), CNRS-INPG-UJF, 1130 rue de la piscine, BP 75, 38402 St. Martin d' Heres Cedex (France)], E-mail: sabine.szunerits@lepmi.inpg.fr; Rich, Sami Abou [Laboratoire de Genie des Procedes d' Interaction de Fluides Reactifs-Materiaux U.S.T.L., Cite Scientifique, 59655 Villeneuve d' Ascq (France); Coffinier, Yannick [Institut de Recherche Interdisciplinaire (IRI), FRE CNRS 2963, Institut d' lectronique, de Microelectronique et de Nanotechnologie (IEMN), UMR CNRS-8520, Cite Scientifique, Avenue Poincare, BP 60069, 59652 Villeneuve d' Ascq (France); Languille, Marie-Angelique [Unite de Catalyse et de Chimie du Solide, UCCS UMR CNRS-8181, Universite des Sciences et Technologies de Lille, Bat. C3, 59655 Villeneuve d' Ascq (France); Supiot, Philippe [Laboratoire de Genie des Procedes d' Interaction de Fluides Reactifs-Materiaux U.S.T.L., Cite Scientifique, 59655 Villeneuve d' Ascq (France); Boukherroub, Rabah [Institut de Recherche Interdisciplinaire (IRI), FRE CNRS 2963, Institut d' lectronique, de Microelectronique et de Nanotechnologie (IEMN), UMR CNRS-8520, Cite Scientifique, Avenue Poincare, BP 60069, 59652 Villeneuve d' Ascq (France)], E-mail: rabah.boukherroub@iemn.univ-lille1.fr

2008-04-20

The paper reports on the preparation and characterization of organosilicon thin polymer films deposited on glass slides coated with 5 nm adhesion layer of titanium and 50 nm of gold. The polymer was obtained by the decomposition of 1,1,3,3-tetramethyldisiloxane precursor (TMDSO) premixed with oxygen induced in a N{sub 2} plasma afterglow using remote plasma-enhanced chemical vapor deposition (PECVD) technique. The film thickness was controlled by laser interferometry and was 9 nm. The chemical stability of the gold substrate coated with the organosilicon polymer film (p-TMDSO) was studied in different acidic and basic solutions (pH 1-14). While the gold/polymer interface showed a high stability in acidic media, the film was almost completely removed in basic solutions. The resulting surfaces were characterized using atomic force microscopy (AFM), X-ray photoelectron spectroscopy (XPS), water contact angle measurements, cyclic voltammetry, and surface plasmon resonance (SPR)

Preparation and characterization of thin organosilicon films deposited on SPR chip

International Nuclear Information System (INIS)

Szunerits, Sabine; Rich, Sami Abou; Coffinier, Yannick; Languille, Marie-Angelique; Supiot, Philippe; Boukherroub, Rabah

2008-01-01

The paper reports on the preparation and characterization of organosilicon thin polymer films deposited on glass slides coated with 5 nm adhesion layer of titanium and 50 nm of gold. The polymer was obtained by the decomposition of 1,1,3,3-tetramethyldisiloxane precursor (TMDSO) premixed with oxygen induced in a N 2 plasma afterglow using remote plasma-enhanced chemical vapor deposition (PECVD) technique. The film thickness was controlled by laser interferometry and was 9 nm. The chemical stability of the gold substrate coated with the organosilicon polymer film (p-TMDSO) was studied in different acidic and basic solutions (pH 1-14). While the gold/polymer interface showed a high stability in acidic media, the film was almost completely removed in basic solutions. The resulting surfaces were characterized using atomic force microscopy (AFM), X-ray photoelectron spectroscopy (XPS), water contact angle measurements, cyclic voltammetry, and surface plasmon resonance (SPR)

Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor

International Nuclear Information System (INIS)

Garcia Martinez, H.

1999-01-01

In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 L ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

Label-free SPR detection of gluten peptides in urine for non-invasive celiac disease follow-up.

Science.gov (United States)

Soler, Maria; Estevez, M-Carmen; Moreno, Maria de Lourdes; Cebolla, Angel; Lechuga, Laura M

2016-05-15

Motivated by the necessity of new and efficient methods for dietary gluten control of celiac patients, we have developed a simple and highly sensitive SPR biosensor for the detection of gluten peptides in urine. The sensing methodology enables rapid and label-free quantification of the gluten immunogenic peptides (GIP) by using G12 mAb. The overall performance of the biosensor has been in-depth optimized and evaluated in terms of sensitivity, selectivity and reproducibility, reaching a limit of detection of 0.33 ng mL(-1). Besides, the robustness and stability of the methodology permit the continuous use of the biosensor for more than 100 cycles with excellent repeatability. Special efforts have been focused on preventing and minimizing possible interferences coming from urine matrix enabling a direct analysis in this fluid without requiring extraction or purification procedures. Our SPR biosensor has proven to detect and identify gluten consumption by evaluating urine samples from healthy and celiac individuals with different dietary gluten conditions. This novel biosensor methodology represents a novel approach to quantify the digested gluten peptides in human urine with outstanding sensitivity in a rapid and non-invasive manner. Our technique should be considered as a promising opportunity to develop Point-of-Care (POC) devices for an efficient, simple and accurate gluten free diet (GFD) monitoring as well as therapy follow-up of celiac disease patients. Copyright © 2015 Elsevier B.V. All rights reserved.

Assaying Auxin Receptor Activity Using SPR Assays with F-Box Proteins and Aux/IAA Degrons.

Science.gov (United States)

Quareshy, Mussa; Uzunova, Veselina; Prusinska, Justyna M; Napier, Richard M

2017-01-01

The identification of TIR1 as an auxin receptor combined with advanced biophysical instrumentation has led to the development of real-time activity assays for auxins. Traditionally, molecules have been assessed for auxinic activity using bioassays, and agrochemical compound discovery continues to be based on "spray and pray" technologies. Here, we describe the methodology behind an SPR-based assay that uses TIR1 and related F-box proteins with surface plasmon resonance spectrometry for rapid compound screening. In addition, methods for collecting kinetic binding data and data processing are given so that they may support programs for rational design of novel auxin ligands.

Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

International Nuclear Information System (INIS)

Viais J, J.

1994-01-01

Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

International Nuclear Information System (INIS)

Su, T.M.

1982-10-01

Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report

Development of a software for the control of the quality management system of the TRIGA-Mark III reactor

International Nuclear Information System (INIS)

Herrera A, E.; Hernandez, L.V.; Hernandez, J.A.

2006-01-01

The quality has not only become one of the essential requirements of the product but rather at the present time it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

Louisiana SIP: LAC 33:III Ch 21 Subchap J, 2147--Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 1998-02-02 (LAc74) to more..

Science.gov (United States)

Louisiana SIP: LAC 33:III Ch 21 Subchap J, 2147--Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 1998-02-02 (LAc74) more...

Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

International Nuclear Information System (INIS)

Barranco R, F.

2015-01-01

In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

International Nuclear Information System (INIS)

Hernandez G, J.

2012-10-01

Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ; Diseno, construccion e implementacion de dos circuitos redundantes de la logica de actuacion del sistema de proteccion de la nueva consola de control del reactor Triga Mark III del ININ

Energy Technology Data Exchange (ETDEWEB)

Celestino M, E.

2016-07-01

The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10. Final report

International Nuclear Information System (INIS)

Fields, M.B.; Kudrick, J.A.

1984-08-01

This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases and conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff approved acceptance criteria for LOCA-related hydrodynamic loads are provided in an appendix

Behavior of exposed human lymphocytes to a neutron beam of the Reactor TRIGA Mark III

International Nuclear Information System (INIS)

Carbajal R, M. I.; Arceo M, C.; Aguilar H, F.; Guerrero C, C.

2012-10-01

The living beings are permanently exposed to radiations of natural origin: cosmic and geologic, as well as the artificial radiations that come from sources elaborated by the man. The artificial sources have an important use in the medical area. Particularly has been increased the neutrons use due to the effectiveness that they have to damage the cells with regard to other radiation types. The biological indicator of exposition to ionizing radiation more reliable is the chromosomal aberrations study, specifically the dicentrics in human lymphocytes. This test allows, establishing the exposition dose in function of the damage quantity. The dicentrics have a behavior in function of the dose. The calibration curve that describes this behavior is specific for each type of ionizing radiation. In the year 2006 beginning was given to the expositions of human lymphocytes to a neutron beam generated in the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico. Up to 2008 the response dose curve comprised an interval of exposition time of up to 30 minutes. Moreover, the interval between 10 an 20 minutes is included, since was observed that this last is indispensable for the adjustment waited in a lineal model. (Author)

SIMMER-III applications to fuel-coolant interactions

Energy Technology Data Exchange (ETDEWEB)

Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1998-01-01

The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

Perspective of nuclear energy and advanced reactors

International Nuclear Information System (INIS)

Lopez Jimenez, J.; Cobian, J.

2007-01-01

Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

RA reactor safety analysis I-III, Part III - Environmental effect of the maximum credible accident; Analiza sigurnosti rada Reaktora RA I-III, III deo - Posledica maksimalno moguceg akcidenta na okolinu reaktora

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

The objective of the maximum credible accident analysis was to determine the integral radiation doses in the vicinity of the reactor and in the environment. In case of RA reactor the maximum credible accident, meaning release of the fission products, would be caused by fuel elements meltdown. This analysis includes the following calculation results: activity of the fission products, volatility of the fission products, concentration of radioactive materials in the air, analysis of the accident environmental effects.

N reactor individual risk comparison to quantitative nuclear safety goals

International Nuclear Information System (INIS)

Wang, O.S.; Rainey, T.E.; Zentner, M.D.

1990-01-01

A full-scope level III probabilistic risk assessment (PRA) has been completed for N reactor, a US Department of Energy (DOE) production reactor located on the Hanford Reservation in the state of Washington. Sandia National Laboratories (SNL) provided the technical leadership for this work, using the state-of-the-art NUREG-1150 methodology developed for the US Nuclear Regulatory Commission (NRC). The main objectives of this effort were to assess the risks to the public and to the on-site workers posed by the operation of N reactor, to identify changes to the plant that could reduce the overall risk, and to compare those risks to the proposed NRC and DOE quantitative safety goals. This paper presents the methodology adopted by Westinghouse Hanford Company (WHC) and SNL for individual health risk evaluation, its results, and a comparison to the NRC safety objectives and the DOE nuclear safety guidelines. The N reactor results, are also compared with the five NUREG-1150 nuclear plants. Only internal events are compared here because external events are not yet reported in the current draft NUREG-1150. This is the first full-scope level III PRA study with a detailed quantitative safety goal comparison performed for DOE production reactors

The UK MK III GCR experimental physics programme at AEE Winfrith

Energy Technology Data Exchange (ETDEWEB)

Johnstone, I

1972-06-15

The UK programme of reactor physics experiments in support of the Mk III GCR project started in 1968/69 and has now reached its third main phase. The overall programme is broadly summarised in this report.

A comparison of FEMAXI-III code calculations with irradiation experiments

International Nuclear Information System (INIS)

Ito, K.; Sogame, M.; Ichikawa, M.; Nakajima, T.

1981-01-01

The FEMAXI-III code calculations were compared with in-pile diameter measurements in the Halden Boiling Water Reactor, in order to check the ability to analyse the pellet-cladding mechanical interaction. The results showed generally good agreement between calculations and measurements. The Studsvik INTER-RAMP Experiments were also analysed to examine the predictability of fuel rod failures. Good agreement was obtained between calculated and measured fission gas x release. The threshold stress to cause failure was estimated by means of FEMAXI-III. (author)

MITR-III: Upgrade and relicensing studies for the MIT Research Reactor. Second annual report

International Nuclear Information System (INIS)

Trosman, H.G.; Lanning, D.D.; Harling, O.K.

1994-08-01

The current operating license of the MIT research reactor will expire on May 7, 1996 or possibly a few years later if the US Nuclear Regulatory Commission agrees that the license period can start with the date of initial reactor operation. Driven by the imminent expiration of the operating license, a team of nuclear engineering staff and students have begun a study of the future options for the MIT Research Reactor. These options have included the range from a major rebuilding of the reactor to its decommissioning. This document reports the results of a two year intensive activity which has been supported by a $148,000 grant from the USDOE contract Number DEFG0293ER75859, approximately $100,000 of internal MIT funds and Nuclear Engineering Department graduate student fellowships as well as assistance from international visiting scientists and engineers

Support to the elaboration of the engineering of detail, configuration and programming of the control system of heat removal of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Diaz G, C. A.

2016-01-01

Nowadays, the peaceful and responsible use of nuclear energy in Mexico is of great importance and contributes to economic, social, scientist and technologic development in the country, highlighting the Instituto Nacional de Investigaciones Nucleares (ININ) and the Nuclear Power Plant of Laguna Verde as one of the most important dependences. Among the main facilities and laboratories of ININ is the Nuclear Research Reactor TRIGA Mark III, this is a pool type reactor with mobile core, cooled and moderated by light water and a flow of 1013 n/cm"2/sec. Due to the technological obsolescence is a growing problem that threatens the information, operation and/or efficacy of elements of control and safety systems of the reactor, these must be changed each time more frequently. In the modernization of reactor was used a Modicon M340 programmable logic control (PLC) and a Twido PLC for the control of heat removal system (Primary Cooling System (PCS) and Secondary Cooling System (SCS) respectively), this because the PLC has proven to be safe and effective devices, addition to reduce the wiring elements and increase the possibilities of performance and design of the digital control console. This document shows and describes the elements of heat removal system (PCS and SCS), and the signals and signal types that such items send or received by the PLC, likewise, is indicated the methodology used to develop the applications for the control of the Primary Cooling System and Secondary Cooling System, beginning with the PLC design, the development of PLC plans and the control logic, and finally, the simulation and debugging of applications on Unity Pro and Twido Suite. All this in compliance with the safety standards to nuclear research reactors (NS-R-4), the rules of industrial programming (IEC 61131-3), and the reactor operating limits postulated in the safety report and the software assurance system used in the ININ. (Author)

Statistical optimization of synthesis procedure and characterization of europium (III) molybdate nano-plates

Energy Technology Data Exchange (ETDEWEB)

Pourmortazavi, Seied Mahdi [Malek Ashtar University of Technology, Faculty of Material and Manufacturing Technologies, P. O. Box 16765-3454, Tehran (Iran, Islamic Republic of); Rahimi-Nasrabadi, Mehdi [Imam Hossein University, Nano Science Center, Tehran (Iran, Islamic Republic of); Fazli, Yousef [Islamic Azad University, Department of Chemistry, Faculty of Science, Arak Branch, Arak (Iran, Islamic Republic of); Mohammad-Zadeh, Mohammad [Sabzevar University of Medical Sciences, Department of Physiology and Pharmacology, School of Medicine, Sabzevar (Iran, Islamic Republic of)

2015-06-15

Europium (III) molybdate nano-plates were synthesized in this work via chemical precipitation route involving adding of europium (III) ion solution to the aqueous solution of molybdate reagent. Effects of some reaction variables such as concentrations of europium and molybdate ions, flow rate of europium reagent, and reactor temperature on the diameter of the synthesized europium (III) molybdate nano-plates were experimentally investigated by orthogonal array design. The results showed that the size of europium (III) molybdate nano-plates can be optimized by adjusting the concentrations of europium (III) and molybdate ions, as well as the reactional temperature. Europium (III) molybdate nano-plates prepared under the optimum conditions were characterized by X-ray powder diffraction, scanning electron microscopy, and Fourier transform infrared spectroscopy. (orig.)

Nuclear data and multigroup methods in fast reactor calculations

International Nuclear Information System (INIS)

Gur, Y.

1975-03-01

The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

Základy fúzní energetiky III. – Reaktorové technologie

Czech Academy of Sciences Publication Activity Database

Entler, Slavomír; Mlynář, Jan; Dostál, V.

Srpen (2016), č. článku 14587. ISSN 1801-4399 Institutional support: RVO:61389021 Keywords : Nuclear fusion * reactor technologies * plasma Subject RIV: JF - Nuclear Energetics http://energetika.tzb-info.cz/elektroenergetika/14587-zaklady-fuzni-energetiky-iii-reaktorove-technologie

The near-threshold high R-ratio fatigue crack growth characteristics of SA508 cl III reactor pressure vessel steel

International Nuclear Information System (INIS)

Achilles, R.D.; Bulloch, J.H.

1989-01-01

This paper describes the effect of frequency and environment on the near-threshold fatigue crack growth behaviour of SA508 cl III reactor pressure vessel (RPV) steel. The study has shown that in the near-threshold regime microstructure and environment markedly affect fatigue crack growth behaviour. In an aqueous environment, fatigue crack growth behaviour became even more sensitive to microstructure, and the fatigue crack growth rate increased by a factor of four in the case of the 3Hz test, while that for the 0.3Hz test was increased by a factor of approximately sixteen. This environmental enhancement manifested itself in the form of intergranular failure. For the 0.3Hz test the percentage intergranular failure decreased from 18% to <1% with an increase in ΔK level. The transition from microstructure-sensitive to microstructure-insensitive occurs when the cyclic plastic zone size is of the order of the prior austenite grain size. (author)

Simultaneous Detection of Escherichia coli O157:H7, Salmonella enteritidis, and Listeria monocytogenes at a Very Low Level Using Simultaneous Enrichment Broth and Multichannel SPR Biosensor.

Science.gov (United States)

Zhang, Xiaoguang; Tsuji, Sachiko; Kitaoka, Hayato; Kobayashi, Hiroshi; Tamai, Mitsuru; Honjoh, Ken-Ichi; Miyamoto, Takahisa

2017-10-01

Detection of foodborne pathogens at very low levels is still a challenge. A custom-built multichannel surface plasmon resonance (SPR) biosensor and simultaneous enrichment broth (SEB) were used to develop a simultaneous detection method for 3 important foodborne pathogens, Escherichia coli O157:H7 (O157:H7), Salmonella enteritidis, and Listeria monocytogenes, at a very low level. These 3 foodborne pathogens at a very low level (14, 6, and 28 CFU/25 g (mL) for O157:H7, S. enteritidis, and L. monocytogenes, respectively) were inoculated in SEB and incubated at 37 ˚C for 24 h. Sample prepared from the simultaneous enrichment culture was analyzed using the multichannel SPR biosensor and sensor chip immobilized with polyclonal antibodies specific to each of the target pathogens. O157:H7, S. enteritidis, and L. monocytogenes in chicken were detected simultaneously at an inoculum dose of 14, 6, and 28 CFU/25 g, respectively. Our method using a custom-built multichannel SPR biosensor and enrichment in SEB is expected as a rapid and simultaneous detection method for low levels of O157:H7, S. enteritidis, and L. monocytogenes in food. Our method is expected as a rapid and simultaneous detection method for pathogens at very low levels. It has great potential for safety control of food and microbiological detection applications. © 2017 Institute of Food Technologists®.

ROP design for Enhanced CANDU 6 reactor

Energy Technology Data Exchange (ETDEWEB)

Hu, J.; Scherbakova, D; Kastanya, D.; Ovanes, M. [Candu Energy Inc., Mississauga, Ontario (Canada)

2011-07-01

The Enhanced CANDU 6 (EC6) nuclear power plant is a mid-sized pressurized heavy water reactor design, based on the highly successful CANDU 6 (C6) family of power plants, upgraded to meet today's Canadian and international safety requirements and to satisfy Generation III expectations. The EC6 reactor is equipped with two independent Regional Overpower Protection (ROP) systems to prevent overpowers in the reactor fuel. The ROP system design, retaining the traditional C6 methodology, is determined to cover the End-of-Life (EOL) reactor core condition since the reactor operating/thermal margin gradually decreases as plant equipment ages. Several design changes have been incorporated into the reference C6 plant to mitigate the ageing effect on the ROP trip margin. This paper outlines the basis for the EC6 ROP physics design and presents the ROP related improvements made in the EC6 design to ensure that full power operation is not limited by the ROP throughout the entire life of the reactor. (author)

Japan: The Experimental Fast Reactor JOYO. Profile 12

International Nuclear Information System (INIS)

2017-01-01

The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

Investigation of SnSPR1, a novel and abundant surface protein of Sarcocystis neurona merozoites.

Science.gov (United States)

Zhang, Deqing; Howe, Daniel K

2008-04-15

An expressed sequence tag (EST) sequencing project has produced over 15,000 partial cDNA sequences from the equine pathogen Sarcocystis neurona. While many of the sequences are clear homologues of previously characterized genes, a significant number of the S. neurona ESTs do not exhibit similarity to anything in the extensive sequence databases that have been generated. In an effort to characterize parasite proteins that are novel to S. neurona, a seemingly unique gene was selected for further investigation based on its abundant representation in the collection of ESTs and the predicted presence of a signal peptide and glycolipid anchor addition on the encoded protein. The gene was expressed in E. coli, and monospecific polyclonal antiserum against the recombinant protein was produced by immunization of a rabbit. Characterization of the native protein in S. neurona merozoites and schizonts revealed that it is a low molecular weight surface protein that is expressed throughout intracellular development of the parasite. The protein was designated Surface Protein 1 (SPR1) to reflect its display on the outer surface of merozoites and to distinguish it from the ubiquitous SAG/SRS surface antigens of the heteroxenous Coccidia. Interestingly, infection assays in the presence of the polyclonal antiserum suggested that SnSPR1 plays some role in attachment and/or invasion of host cells by S. neurona merozoites. The work described herein represents a general template for selecting and characterizing the various unidentified gene sequences that are plentiful in the EST databases for S. neurona and other apicomplexans. Furthermore, this study illustrates the value of investigating these novel sequences since it can offer new candidates for diagnostic or vaccine development while also providing greater insight into the biology of these parasites.

Planning for the decommissioning of a research reactor

International Nuclear Information System (INIS)

Dodson, W.J.; Isakari, H.H.; Munro, J.F.; Lim, T.H.; Denton, M.M.; Vernig, P.G.

1988-01-01

This paper describes the steps that must be taken and the uncertainties and potential pitfalls that can be encountered in decommissioning a research reactor, whether owned by private industry, a university, or a government agency. The paper is based on the experience in preparing for decommissioning the TRIGA Mark III Berkeley Research Reactor (BRR). Six topics of interest to an owner-operator are addressed: task and schedule planning, decommissioning organization, cost estimating, health and safety considerations, waste management, and regulatory concerns

Fast wave current drive in reactor scale tokamaks

International Nuclear Information System (INIS)

Moreau, D.

1992-01-01

The IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks, hosted by the Commissariat a l'Energie Atomique (CEA), Departement de Recherches sur la Fusion Controlee (Centres d'Etudes de Cadarache, under the Euratom-CEA Association for fusion) aimed at discussing the physics and the efficiency of non-inductive current drive by fast waves. Relevance to reactor size tokamaks and comparison between theory and experiment were emphasized. The following topics are described in the summary report: (i) theory and modelling of radiofrequency current drive (theory, full wave modelling, ray tracing and Fokker-Planck calculations, helicity injection and ponderomotive effects, and alternative radio-frequency current drive effects), (ii) present experiments, (iii) reactor applications (reactor scenarios including fast wave current drive; and fast wave current drive antennas); (iv) discussion and summary. 32 refs

Development of an alanine dosimeter for gamma dosimetry in mixed environments

International Nuclear Information System (INIS)

Vehar, D.W.; Griffin, P.J.

1992-01-01

L-αa-Alanine, a nontoxic polycrystalline amino acid, has been investigated for use in high-precision, high-level absorbed-dose measurements in mixed neutron/photon environments such as research and test reactors. The technique is based on the use of electron paramagnetic resonance (EPR) spectroscopy to determine the extent of free radical production in a sample exposed to ionizing radiation, and has been successfully used for photon absorbed-dose measurements at levels exceeding 10 5 Gy with high measurement precision. Application of the technique to mixed environments requires knowledge of the energy-dependent response of the dosimeter for both photons and neutrons. Determination of the dosimeter response to photons is accomplished by irradiations in 60 Co and bremsstrahlung sources and by calculations of energy-dependent photon kerma. Neutron response is determined by calculations in conjunction with CaF 2 :Mn thermoluminescence dosimeters and by calculations of energy-dependent neutron kerma. Several neutron environments are used, including the ACRR and SPR-III reactors

Observations on vapor pressure in SPR caverns : sources.

Energy Technology Data Exchange (ETDEWEB)

Munson, Darrell Eugene

2010-05-01

The oil of the Strategic Petroleum Reserve (SPR) represents a national response to any potential emergency or intentional restriction of crude oil supply to this country, and conforms to International Agreements to maintain such a reserve. As assurance this reserve oil will be available in a timely manner should a restriction in supply occur, the oil of the reserve must meet certain transportation criteria. The transportation criteria require that the oil does not evolve dangerous gas, either explosive or toxic, while in the process of transport to, or storage at, the destination facility. This requirement can be a challenge because the stored oil can acquire dissolved gases while in the SPR. There have been a series of reports analyzing in exceptional detail the reasons for the increases, or regains, in gas content; however, there remains some uncertainty in these explanations and an inability to predict why the regains occur. Where the regains are prohibitive and exceed the criteria, the oil must undergo degasification, where excess portions of the volatile gas are removed. There are only two known sources of gas regain, one is the salt dome formation itself which may contain gas inclusions from which gas can be released during oil processing or storage, and the second is increases of the gases release by the volatile components of the crude oil itself during storage, especially if the stored oil undergoes heating or is subject to biological generation processes. In this work, the earlier analyses are reexamined and significant alterations in conclusions are proposed. The alterations are based on how the fluid exchanges of brine and oil uptake gas released from domal salt during solutioning, and thereafter, during further exchanges of fluids. Transparency of the brine/oil interface and the transfer of gas across this interface remains an important unanswered question. The contribution from creep induced damage releasing gas from the salt surrounding the cavern is

Protein-Flavonoid Interaction Studies by a Taylor Dispersion Surface Plasmon Resonance (SPR Technique: A Novel Method to Assess Biomolecular Interactions

Directory of Open Access Journals (Sweden)

Preejith P. Vachali

2016-02-01

Full Text Available Flavonoids are common polyphenolic compounds widely distributed in fruits and vegetables. These pigments have important pharmacological relevance because emerging research suggests possible anti-cancer and anti-inflammatory properties as well other beneficial health effects. These compounds are relatively hydrophobic molecules, suggesting the role of blood transport proteins in their delivery to tissues. In this study, we assess the binding interactions of four flavonoids (kaempferol, luteolin, quercetin, and resveratrol with human serum albumin (HSA, the most abundant protein in the blood, and with glutathione S-transferase pi isoform-1 (GSTP1, an enzyme with well-characterized hydrophobic binding sites that plays an important role in detoxification of xenobiotics with reduced glutathione, using a novel Taylor dispersion surface plasmon resonance (SPR technique. For the first time, HSA sites revealed a high-affinity binding site for flavonoid interactions. Out of the four flavonoids that we examined, quercetin and kaempferol showed the strongest equilibrium binding affinities (KD of 63 ± 0.03 nM and 37 ± 0.07 nM, respectively. GSTP1 displayed lower affinities in the micromolar range towards all of the flavonoids tested. The interactions of flavonoids with HSA and GSTP1 were studied successfully using this novel SPR assay method. The new method is compatible with both kinetic and equilibrium analyses.

Bioremoval of trivalent chromium using Bacillus biofilms through continuous flow reactor

International Nuclear Information System (INIS)

Sundar, K.; Sadiq, I. Mohammed; Mukherjee, Amitava; Chandrasekaran, N.

2011-01-01

Highlights: ► Effective bioremoval of Cr(III) using bacterial biofilms. ► Simplified bioreactor was fabricated for the biofilm development and Cr(III) removal. ► Economically feasible substrate like coarse sand and pebbles were used. - Abstract: Present study deals with the applicability of bacterial biofilms for the bioremoval of trivalent chromium from tannery effluents. A continuous flow reactor was designed for the development of biofilms on different substrates like glass beads, pebbles and coarse sand. The parameters for the continuous flow reactor were 20 ml/min flow rate at 30 °C, pH4. Biofilm biomass on the substrates was in the following sequence: coarse sand > pebbles > glass beads (4.8 × 10 7 , 4.5 × 10 7 and 3.5 × 10 5 CFU/cm 2 ), which was confirmed by CLSM. Biofilms developed using consortium of Bacillus subtilis and Bacillus cereus on coarse sand had more surface area and was able to remove 98% of Cr(III), SEM-EDX proved 92.60% Cr(III) adsorption on biofilms supported by coarse sand. Utilization of Bacillus biofilms for effective bioremoval of Cr(III) from chrome tanning effluent could be a better option for tannery industry, especially during post chrome tanning operation.

Facility management - efektivní správa stavebních objektů

OpenAIRE

Helekalová, Denisa

2013-01-01

Diplomová práce „ Facility management – efektivní správa stavebních objektů“ je zaměřena na stavební objekty a přínos pro ně při využívání služeb facility managementu. Práce se snaží přiblížit, co je facility management kdo ho poskytuje v České republice a jak přispívá k úsporám. Zmíněny jsou také informační technologie využívané ve facility managementu, protože bez těch se v dnešní době téměř žádný obor neobejde. První část je též zaměřena na stavbu, její životní cyklus a náklady s ním spoje...

Operation experience at the TRIGA Mark III reactor FRN within the last two years

International Nuclear Information System (INIS)

Rau, G.

1976-01-01

A survey is given of the history of FRN-reactor operation since the last TRIGA conference. In 1975, the reactor had been operated for 830 MWh and had been pulsed 1050 times. In the first half of 1976, an appreciable Increase of reactor activities took place. In this time interval, the integrated power amounted to 996 MWh and the number of released pulses to 1024. In 1975, an important event was the exchange of the defective rotary specimen rack against a water-operated type. For this purpose, the whole core had to be unloaded and the pool water had to be removed. In this connection, damages at the beam-port weldings had been discovered and had to be repaired immediately. In December 1975, a new core configuration had been set up by adding ten spare fuel elements into the F-resp. G-ring. The gain of excess reactivity amounted to approximately 1,8 $. This however allowed reactor operation without disturbance by Xe-poisoning only for a limited time interval. (author)

Design improvements in TRIGA reactors

International Nuclear Information System (INIS)

Batch, John M.

1970-01-01

There have been many design improvements to TRIGA reactor hardware in the past twelve years. One of the more important and most obvious improvements has been in the area of reactor instrumentation. The low profile, completely transistorized Mark III console was a great step forward in a low maintenance, high reliability instrumentation system. Other design improvements include the lazy susan specimen pickup assembly; the specimen container; an empty stainless steel fuel element which can be filled with samples and can be located anywhere in the core; the flexible fuel handling tool; a new fuel measuring tool design; the shock absorber on the adjustable transient rod drive; new testing and evaluation procedures on the thermocouples and other

Strategic Petroleum Reserve (SPR) additional geologic site characterization studies, Bryan Mound Salt Dome, Texas

Energy Technology Data Exchange (ETDEWEB)

Neal, J.T. [Sandia National Labs., Albuquerque, NM (United States); Magorian, T.R.; Ahmad, S. [Acres International Corp., Amherst, NY (United States)

1994-11-01

This report revises the original report that was published in 1980. Some of the topics covered in the earlier report were provisional and it is now practicable to reexamine them using new or revised geotechnical data and that obtained from SPR cavern operations, which involves 16 new caverns. Revised structure maps and sections show interpretative differences as compared with the 1980 report and more definition in the dome shape and caprock structural contours, especially a major southeast-northwest trending anomalous zone. The original interpretation was of westward tilt of the dome, this revision shows a tilt to the southeast, consistent with other gravity and seismic data. This interpretation refines the evaluation of additional cavern space, by adding more salt buffer and allowing several more caverns. Additional storage space is constrained on this nearly full dome because of low-lying peripheral wetlands, but 60 MMBBL or more of additional volume could be gained in six or more new caverns. Subsidence values at Bryan Mound are among the lowest in the SPR system, averaging about 11 mm/yr (0.4 in/yr), but measurement and interpretation issues persist, as observed values are about the same as survey measurement accuracy. Periodic flooding is a continuing threat because of the coastal proximity and because peripheral portions of the site are at elevations less than 15 ft. This threat may increase slightly as future subsidence lowers the surface, but the amount is apt to be small. Caprock integrity may be affected by structural features, especially the faulting associated with anomalous zones. Injection wells have not been used extensively at Bryan Mound, but could be a practicable solution to future brine disposal needs. Environmental issues center on the areas of low elevation that are below 15 feet above mean sea level: the coastal proximity and lowland environment combined with the potential for flooding create conditions that require continuing surveillance.

Application of the subchannel analysis code COBRA III C for liquid sodium

International Nuclear Information System (INIS)

Nissen, K.L.

1981-01-01

The subchannel-analysis code COBRA III C was developed to gain knowledge of mass flow and temperature distribution in rod bundles of light water reactors. A comparison of experimental results for the temperature distribution in a 19 rod bundle with calculations done by the computer program shows the capability of COBRA III C to handle liquid sodium cooling. The code needs sodium properties as well as changed correlations for turbulent mixing and heat transfer at the rod. (orig.) [de

Decontamination and decommissioning project status of the TRIGA Mark II and III in Korea

International Nuclear Information System (INIS)

Paik, S.T.; Park, S.K.; Chung, K.W.; Chung, U.S.; Jung, K.J.

1999-01-01

TRIGA Mark-II, the first research reactor in Korea, has operated since 1962, and the second one, TRIGA Mark-III since 1972. Both of them had their operation phased out in 1995 due to their lives and operation of the new research reactor, HANARO (High-flux Advanced Neutron Application Reactor) at the Korea Atomic Energy Institute (KAERI) in Taejon. Decontamination and decommissioning (D and D) project of TRIGA Mark-II and Mark-III was started in January 1997 and will be completed in December 2002. The first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of Korea Institute of Nuclear Safety (KINS). Hyundai Engineering Company (HEC) is the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels plc (BNFL) is the technical assisting partner of HEC. The decommissioning plan document was submitted to the Ministry of Since and Technology (MOST) for the decommissioning license in December 1998, and it expecting to be issued a license in mid 1999. The goal of this project is to release the reactor site and buildings as an unrestricted area. This paper summarizes current status and future plan for the D and D project. (author)

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).

Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ

International Nuclear Information System (INIS)

Celestino M, E.

2016-01-01

The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

Use of surface plasmon resonance (SPR) to study the dissociation and polysaccharide binding of casein micelles and caseins.

Science.gov (United States)

Thompson, Abby K; Singh, Harjinder; Dalgleish, Douglas G

2010-11-24

Tests were made to determine whether surface plasmon resonance (SPR) could be used as a technique to study the dissociation properties of bovine casein micelles or of sodium caseinate and the interactions between these protein particles and different polysaccharides. Surfaces of bound micelles or caseinate were made, and the changes in refractive index of these layers were used to define changes in the structures of the chemisorbed material. The technique appears to have some potential for studying details of the dissociation of casein micelles and of the binding of different polysaccharides to caseins.

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (III). The third experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-03-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the third reactor noise experiment using pseudo random binary sequences (PRBS) was performed on September 16, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 70% of reactor power and under a normal 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. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning 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)

Study on the decommissioning of research reactor

Energy Technology Data Exchange (ETDEWEB)

Suh, Doo Hwan; Jun, Kwan Sik; Choi, Yoon Dong; Lee, Tae Yung; Kwon, Sang Woon; Lee, Jong Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-01-01

Currently, KAERI operates TRIGA Mark-II and TRIGA Mark-III research reactors as a general purpose research and training facility. As these are, however, situated at Seoul office site of KAERI which is scheduled to be transferred to KEPCO as well as 30 MW HANARO research reactor which is expected to reach the first criticality in 1995 is under construction at head site of KAERI, decommissioning of TRIGA reactors has become an important topic. The objective of this study is to prepare and present TRIGA facility decontamination and decommissioning plan. Estimation of the radioactive inventory in TRIGA research reactor was carried out by the use of computational method. In addition, summarized in particular were the methodologies associated with decontamination, segmenting processes for activated metallic components, disposition of wastes. Particular consideration in this study was focused available technology applicable to decommissioning of TRIGA research reactor. State-of-the-art summaries of the available technology for decommissioning presented here will serve a useful document for preparations for decommissioning in the future. 6 figs, 41 tabs, 30 refs. (Author).

Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2002-07-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

MHD stability regimes for steady state and pulsed reactors

International Nuclear Information System (INIS)

Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

1994-02-01

A tokamak reactor will operate at the maximum value of β≡2μ 0 /B 2 that is compatible with MHD stability. This value depends upon the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near one, I bs /I p ∼ 1, which constrains the product of the inverse aspect ratio and the plasma poloidal beta to be near unity, ε β p ∼ 1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during the ARIES I, II/IV, and III and the PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements on the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies is also discussed

Magnetohydrodynamic stability regimes for steady state and pulsed reactors

International Nuclear Information System (INIS)

Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

1994-01-01

A tokamak reactor will operate at the maximum value of β≡2μ 0 left angle p right angle /B 2 that is compatible with magnetohydrodynamic (MHD) stability. This value depends on the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near unity, I BS /I P ∼1, which constrains the product of the inverse aspect ratio and the plasma poloidal β to be near unity, arepsilonβ P ∼1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during ARIES I, II/IV, and III and PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements in the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies, is also discussed. ((orig.))

Fluorescence Resonance Energy Transfer of the Tb(III)-Nd(III) Binary System in Molten LiCl-KCl Eutectic Salt

Energy Technology Data Exchange (ETDEWEB)

Kim, B. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yun, J. I. [KAIST, Daejeon (Korea, Republic of)

2015-05-15

The lanthanides act as a neutron poison in nuclear reactor with large neutron absorption cross section. For that reason, very low amount of lanthanides is required in the recovered U/TRU ingot product from pyrochemical process. In view of that, the investigation of thermodynamic properties and chemical behaviors of lanthanides in molten chloride salt are necessary to estimate the performance efficiency of pyrochemical process. However, there are uncertainties about knowledge and understanding of basic mechanisms in pyrochemical process, such as chemical speciation and redox behaviors due to the lack of in-situ monitoring methods for high temperature molten salt. The spectroscopic analysis is one of the probable techniques for in-situ qualitative and quantitative analysis. Recently, a few fluorescence spectroscopic measurements on single lanthanide element in molten LiCl-KCl eutectic have been investigated. The fluorescence intensity and the fluorescence lifetime of Tb(III) were decreased as increasing the concentration of Nd(III), demonstrating collisional quenching between donor ions and acceptor ions. The Forster distance (..0) of Tb(III)-Nd(III) binary system in molten LiCl-KCl eutectic was determined in the specific range of .... (0.1-1.0) and .. (1.387-1.496)

A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

International Nuclear Information System (INIS)

Bartram, B.W.; Dougherty, D.K.

1987-01-01

This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs

Analysis of reactivity insertion accidents in PWR reactors

International Nuclear Information System (INIS)

Camargo, C.T.M.

1978-06-01

A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt

Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part III

International Nuclear Information System (INIS)

1978-09-01

The Summary Report - Part III of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercialization LMFBRs according to national plans, mostly related to technology of fuel fabrication, PHENIX fuel pins testing, heterogeneous cores, in service inspection of fuel elements, regulations and licensing, and related OECD activities. Most of the discussions were related to the existing reactors: BR-10, BN-600, BN-350, BN-1600, RAPSODIE and PHENIX

Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part III

Energy Technology Data Exchange (ETDEWEB)

NONE

1978-09-01

The Summary Report - Part III of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercialization LMFBRs according to national plans, mostly related to technology of fuel fabrication, PHENIX fuel pins testing, heterogeneous cores, in service inspection of fuel elements, regulations and licensing, and related OECD activities. Most of the discussions were related to the existing reactors: BR-10, BN-600, BN-350, BN-1600, RAPSODIE and PHENIX.

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

Bioremoval of trivalent chromium using Bacillus biofilms through continuous flow reactor

Energy Technology Data Exchange (ETDEWEB)

Sundar, K.; Sadiq, I. Mohammed; Mukherjee, Amitava [Centre for Nanobiotechnology, Nano Bio-Medicine Laboratory School of Bio Sciences and Technology VIT University, Vellore - 632014 (India); Chandrasekaran, N., E-mail: nchandrasekaran@vit.ac.in [Centre for Nanobiotechnology, Nano Bio-Medicine Laboratory School of Bio Sciences and Technology VIT University, Vellore - 632014 (India)

2011-11-30

Highlights: Black-Right-Pointing-Pointer Effective bioremoval of Cr(III) using bacterial biofilms. Black-Right-Pointing-Pointer Simplified bioreactor was fabricated for the biofilm development and Cr(III) removal. Black-Right-Pointing-Pointer Economically feasible substrate like coarse sand and pebbles were used. - Abstract: Present study deals with the applicability of bacterial biofilms for the bioremoval of trivalent chromium from tannery effluents. A continuous flow reactor was designed for the development of biofilms on different substrates like glass beads, pebbles and coarse sand. The parameters for the continuous flow reactor were 20 ml/min flow rate at 30 Degree-Sign C, pH4. Biofilm biomass on the substrates was in the following sequence: coarse sand > pebbles > glass beads (4.8 Multiplication-Sign 10{sup 7}, 4.5 Multiplication-Sign 10{sup 7} and 3.5 Multiplication-Sign 10{sup 5} CFU/cm{sup 2}), which was confirmed by CLSM. Biofilms developed using consortium of Bacillus subtilis and Bacillus cereus on coarse sand had more surface area and was able to remove 98% of Cr(III), SEM-EDX proved 92.60% Cr(III) adsorption on biofilms supported by coarse sand. Utilization of Bacillus biofilms for effective bioremoval of Cr(III) from chrome tanning effluent could be a better option for tannery industry, especially during post chrome tanning operation.

Decontamination of soil from the research reactor site

International Nuclear Information System (INIS)

Won, H. Z.; Kim, K. N.; Choi, W. K.; Jeong, J. H.; Oh, W. J.

2002-01-01

The two research reactors (TRIGA MARK II and III) in Korea are to be decommissioned in the near future. When the reactors are completely dismantled, the site may remain contaminated due to the long period of operation. We assume that the site is radioactively contaminated by Co-60. Soils gathered from the research reactor site were artificially contaminated with Co 2+ ion. The desorption characteristics of Co 2+ ion from the soil surface by citric acid solution were investigated. Decontamination performances of citric acid and EDTA on soil stored in the radioactive waste drums was examined. The feasibility test of recycling the citric acid was also performed. We concluded that the radioactive waste volume could be reduced significantly by soil washing with a citric acid solution

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

Science.gov (United States)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

Directory of Open Access Journals (Sweden)

Wagemans Jan

2016-01-01

Full Text Available The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

Tenth annual meeting, Vienna, Austria, 29 March - 1 April 1977. Summary report. Part III

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-11-01

The Summary Report - Part III of the Tenth Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercial development of FBRs according to national plans, mostly related to technology problems of containment design, fuel fabrication, fuel failures, sodium pressure, fuel-sodium interaction, computer codes needed for licensing. Most of the discussions were related to the existing reactors: BN-600, BN-350, BN-1600, BOR-60, RAPSODIE, PHENIX.

Tenth annual meeting, Vienna, Austria, 29 March - 1 April 1977. Summary report. Part III

International Nuclear Information System (INIS)

1977-11-01

The Summary Report - Part III of the Tenth Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercial development of FBRs according to national plans, mostly related to technology problems of containment design, fuel fabrication, fuel failures, sodium pressure, fuel-sodium interaction, computer codes needed for licensing. Most of the discussions were related to the existing reactors: BN-600, BN-350, BN-1600, BOR-60, RAPSODIE, PHENIX

Design and analysis of surface plasmon resonance (SPR) sensor to check the quality of food from adulteration

Science.gov (United States)

Kumar, Manish; Raghuwanshi, Sanjeev Kumar

2018-02-01

In recent years, food safety issues caused by contamination of chemical substances or microbial species have raised a major area of concern to mankind. The conventional chromatography-based methods for detection of chemical are based on human-observation and slow for real-time monitoring. The surface plasmon resonance (SPR) sensors offers the capability of detection of very low concentrations of adulterated chemical and biological agents for real-time by monitoring. Thus, adulterant agent in food gives change in refractive index of pure food result in corresponding phase change. These changes can be detected at the output and can be related to the concentration of the chemical species present at the point.

Modernization of reactor instrumentation for research reactors at Trombay

International Nuclear Information System (INIS)

Darbhe, M.D.; Chaudhuri, H.

1989-01-01

The three research reactors at Trombay, viz., Apsara, Cirus and Zerlina were commissioned in 1956, 1960 and 1961 respectively. The nuclear instrumentation designs were based on the vacuum tube technology, which was prevalent during those days. The effect of component obsolescence of critical components like vacuum tubes, magnetic amplifiers and sensitrol meter relays was strongly felt since early 1970s. Also, the failure rates of the units were observed to show an increasing trend due to ageing and lack of good quality indigenous spares. Hence it was proposed to replace the nuclear instrumentation units for the three reactors, with those employing modern, state of the art solid state devices, keeping indigenous content as high as practicable. The work started in 1977 with the preparations of specifications and the project was scheduled to be completed in 1981. The project was divided into two phases. The Phase I comprising of nuclear channels common to all reactors and Phase II consisting exclusively of regulating system units of Cirus. The salient stages of project progress and completion were: (i) Fabrication and testing of final design prototypes was completed by end of 1982. (ii) Commissioning of new units at Apsara was completed in January 1984. (iii) Commissioning of new units at Cirus was completed in September 1984. An account of experience in all these stages and problems encountered is given. (author). 6 figs

Nitrato-complexes of Y(III), La(III), Ce(III), Pr(III), Nd(III), Sm(III), Gd(III), Tb(III), Dy(III) and Ho(III) with 2-(2'-pyridyl) benzimidazole

Energy Technology Data Exchange (ETDEWEB)

Mishra, A; Singh, M P; Singh, V K

1982-05-01

The nitrato-complexes, (Y(PyBzH)/sub 2/(NO/sub 3/)/sub 2/)NO/sub 3/.H/sub 2/O and Nd, Sm, Gd, Tb, Dy, Ho ; n=1-3, m=0-0.5 ; PyBzh=2-(2 -pyridyl)benzimidazole) are formed on interaction of the ligand with metal nitrates in ethanol. The electrical conductance values (116-129 ohm/sup -1/cm/sup 2/mol/sup -1/) suggest 1:1 electrolyte-nature of the complexes. Magnetic moment values of Ce(2.53 B.M.), Pr(3.62 B.M.), Nd(3.52 B.M.), Sm(1.70 B.M.), Gd(8.06 B.M.), Tb(9.44 B.M.), Dy(10.56 B.M.) and Ho(10.51 B.M.) in the complexes confirm the positive state of the metals. Infrared evidences are obtained for the existance of both coordinated (C/sub 2/v) and uncoordinated (D/sub 3/h) nitrate groups. Electronic absorption spectra of Pr(III)-, Nd(III)-, Sm(III)-, Tb(III)-, Dy(III)- and Ho(III)-complexes have been analysed in the light of LSJ terms.

Nitrato-complexes of Y(III), La(III), Ce(III), Pr(III), Nd(III), Sm(III), Gd(III), Tb(III), Dy(III) and Ho(III) with 2-(2'-pyridyl) benzimidazole

International Nuclear Information System (INIS)

Mishra, A.; Singh, M.P.; Singh, V.K.

1982-01-01

The nitrato-complexes, [Y(PyBzH) 2 (NO 3 ) 2 ]NO 3 .H 2 O and Nd, Sm, Gd, Tb, Dy, Ho ; n=1-3, m=0-0.5 ; PyBzh=2-(2 -pyridyl)benzimidazole] are formed on interaction of the ligand with metal nitrates in ethanol. The electrical conductance values (116-129 ohm -1 cm 2 mol -1 ) suggest 1:1 electrolyte-nature of the complexes. Magnetic moment values of Ce(2.53 B.M.), Pr(3.62 B.M.), Nd(3.52 B.M.), Sm(1.70 B.M.), Gd(8.06 B.M.), Tb(9.44 B.M.), Dy(10.56 B.M.) and Ho(10.51 B.M.) in the complexes confirm the terpositive state of the metals. Infrared evidences are obtained for the existance of both coordinated (C 2 v) and uncoordinated (D 3 h) nitrate groups. Electronic absorption spectra of Pr(III)-, Nd(III)-, Sm(III)-, Tb(III)-, Dy(III)- and Ho(III)-complexes have been analysed in the light of LSJ terms. (author)

EPR by Areva. EPR the 1600+ MWe reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-07-01

This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system.

EPR by Areva. EPR the 1600+ MWe reactor

International Nuclear Information System (INIS)

2008-01-01

This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system

Operation and maintenance of the RA reactor in 1964, I-II, Part I; Pogon i odrzavanje reaktora RA u 1964. godini, I-II, I Deo

Energy Technology Data Exchange (ETDEWEB)

Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1964-12-15

During 1964, the Reactor as operated about 20 days each months at nominal power of 6.5 MW, 5 days at lower power levels and 5 days were used for maintenance. Total production was 27930 MWh which is 11.7% higher than the planned value. Fuel exchange was done 3 times during this period, 98 spent fuel channels were exchanged. In addition to routine maintenance of reactor components and instruments a series of analyses of heavy water and helium were done. Special attention was devoted to corrosion analyses of the reactor materials because of the heavy water system was refurbished decontaminated in 1963. Utilization of the experimental space in the reactor was better that previously. 546 samples were irradiated till the end of November, of which 443 for users from the Institute. Specific irradiations in the fast neutron flux were done in six VISA-2 channels in the core.

CESEC III code conversion from Apollo to HP9000

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-01-01

CESEC code is a computer program used to analyze transient behaviour of reactor coolant systems of nuclear power plants. CESEC III is an extension of original CESEC code in order to apply wide range of accident analysis including ATWS model. Major parameters during the transients are calculated by CESEC. This report firstly describes detailed work carried out for installation of CESEC III on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also described in relation to installation of CESECIII on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

CESEC III code conversion from Apollo to HP9000

Energy Technology Data Exchange (ETDEWEB)

Lee, Hae Cho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-01-01

CESEC code is a computer program used to analyze transient behaviour of reactor coolant systems of nuclear power plants. CESEC III is an extension of original CESEC code in order to apply wide range of accident analysis including ATWS model. Major parameters during the transients are calculated by CESEC. This report firstly describes detailed work carried out for installation of CESEC III on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also described in relation to installation of CESECIII on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new.

Possible incorporation of a dee-shaped vacuum vessel in Doublet III

International Nuclear Information System (INIS)

Davis, L.; Rawls, J.M.

1979-11-01

The design of Doublet III allows relative straightforward incorporation of any of a number of possible dee-configuration vacuum vessels that can serve as relevant size tests of reactor regime devices. Configurations simulating those of JET, ETF and INTOR with plasma areas larger than TFTR can be attained with significant physics parameter results. Such modifications to Doublet III could be incorporated into planned upgrade activites with operations beginning in 1984, early enough to influence the designs of ETF and INTOR and test the scaling laws, poloidal coil system, and impurity control systems proposed for these ignition devices

Research reactor's role in Korea

International Nuclear Information System (INIS)

Choi, C-O.

1995-01-01

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960's in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs

Behavior of exposed human lymphocytes to a neutron beam of the reactor TRIGA Mark III; Comportamiento de linfocitos humanos expuestos a un haz de neutrones del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Carbajal R, M. I.

2012-07-01

Excessive exposure to ionizing radiation occurs in people who require radiation treatment, also in those for work can come to receive doses above the permitted levels. A third possibility of exposure is the release of radioactive material in which the general population is affected. Most of the time the exhibition is partial and only rarely occurs throughout the body. For various reasons, situations arise where it is impossible to determine by conventional physical methods, the amount of radiation you were exposed to the affected person and in these cases where the option to follow is the Biological Dosimetry, where the analysis of chromosomes dicentrics is used to estimate the dose of ionizing radiation exposure. A calibration curve is generated from in vitro analysis of dicentric chromosome, which are found in human lymphocytes, treated with different types and doses of radiation. The dicentric is formed from two lesions, one on each chromosome and their union results in a structure having two centromeres, acentric fragment with her for the union of several chromosomes leads to more complex structures as tri-centric s, tetra or penta-centric s, which have the same origin. The dose-response curve is estimated by observing the frequency of dicentrics and extrapolated to a dose-effect curve previously established, for which it is necessary that each lab has its own calibration curves, taking into account that for a Let low radiation, dose-effect curve follows a linear-quadratic model Y=C + {alpha}D + {beta}D. The production of dicentric chromosomes with a high Let, was studied using a beam of neutrons generated in the reactor TRIGA Mark III with an average energy of 1 MeV, adjusting the linear model Y={alpha}D. The dose-response relationship is established in blood samples from the same donor, the coefficient {alpha} of the dose-response is Y = (0.3692 {+-} 0.011 * D), also shows that saturation is reached in system 4 Gy. (Author)

A way to limit the corrosion in the Molten Salt Reactor concept: the salt redox potential control

International Nuclear Information System (INIS)

Gibilaro, M.; Massot, L.; Chamelot, P.

2015-01-01

The possibility of controlling the salt redox potential thanks to a redox buffer in the Molten Salt Fast Reactor was investigated, the goal was to limit the oxidation of the reactor structural material. Tests were performed in LiF-CaF 2 at 850 °C on two different redox couples to fix the salt potential, Eu(III)/Eu(II) and U(IV)/U(III), where the first one was used as inactive system to validate the methodology to be applied on the uranium system. A metallic reducing agent (Gd plate for Eu, and U plate for U system) was inserted in the salt, leading to a spontaneous reaction: Eu(III) and U(IV) were then reduced. Eu(III) was fully converted into Eu(II) with metallic Gd, validating the approach. On the U system, the U(IV)/U(III) ratio has to be set between 10 and 100 to limit the core material oxidation: addition of metallic U decreased the concentration ratio from the infinite to 1, showing the feasibility of the salt redox potential control with the U system

Cogeneration using a nuclear reactor to generate process heat

International Nuclear Information System (INIS)

Alonso, Gustavo; Ramirez, Ramon

2009-01-01

Some of the new nuclear reactor technologies (Generation III+) are claiming the production of process heat as an additional value to electricity generation. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product. The current study assess the likeliness of generate process heat from a Pebble Bed Modular Reactor to be used for a refinery showing different plant balance and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor and also the challenges that this option has. (author)

Damage by radiation in structural materials of BWR reactor vessels

International Nuclear Information System (INIS)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2002-01-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B.

2003-01-01

The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

Kinetic analysis of inhibition of glucoamylase and active site mutants via chemoselective oxime immobilization of acarbose on SPR chip surfaces

DEFF Research Database (Denmark)

Sauer, Jørgen; Abou Hachem, Maher; Svensson, Birte

2013-01-01

shown that at pH 7.0 the association and dissociation rate constants for the acarbose-glucoamylase interaction are 104M−1s−1 and 103s−1, respectively, and that the conformational change to a tight enzyme–inhibitor complex affects the dissociation rate constant by a factor of 102s−1. Additionally......, the acarbose-presenting SPR surfaces could be used as a glucoamylase sensor that allowed rapid, label-free affinity screening of small carbohydrate-based inhibitors in solution, which is otherwise difficult with immobilized enzymes or other proteins....

Nuclear reactor control with fuzzy logic approaches - strengths, weakness, opportunities, and threats

International Nuclear Information System (INIS)

Ruan, Da

2004-01-01

As part of the special track on 'Lessons learned from computational intelligence in nuclear applications' at the forthcoming FLINS 2004 conference on Applied Computational Intelligence (Blankenberge, Belgium, September 1-3, 2004), research experiences on fuzzy logic techniques in applications of nuclear reactor control operation are critically reviewed in this presentation. Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined thought a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK-CEN) and the Mexican Nuclear Centre (ININ) on the fuzzy logic control for nuclear reactor control project under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (Author)

Influence of weak magnetic field and tartrate on the oxidation and sequestration of Sb(III) by zerovalent iron: Batch and semi-continuous flow study.

Science.gov (United States)

Fan, Peng; Sun, Yuankui; Qiao, Junlian; Lo, Irene M C; Guan, Xiaohong

2018-02-05

The influence of weak magnetic field (WMF) and tartrate on the oxidation and sequestration of Sb(III) by zerovalent iron (ZVI) was investigated with batch and semi-continuous reactors. The species analysis of antinomy in aqueous solution and solid precipitates implied that both Sb(III) adsorption preceding its conversion to Sb(V) in solid phase and Sb(III) oxidation to Sb(V) preceding its adsorption in aqueous phase occurred in the process of Sb(III) sequestration by ZVI. The application of WMF greatly increased the rate constants of Sb tot (total Sb) and Sb(III) disappearance during Sb(III)-tartrate and uncomplexed-Sb(III) sequestration by ZVI. The enhancing effect of WMF was primarily due to the accelerated ZVI corrosion in the presence of WMF, as evidenced by the influence of WMF on the change of solution and solid properties with reaction. However, tartrate greatly retarded Sb removal by ZVI. It was because tartrate inhibited ZVI corrosion, competed with Sb(III) and Sb(V) for the active surface sites, increased the negative surface charge of the generated iron (hydr)oxides due to its adsorption, and formed soluble complexes with Fe(III). The positive effect of WMF on Sb(III)-tartrate and uncomplexed-Sb(III) removal by ZVI was also verified with a magnetic semi-continuous reactor. Copyright © 2017 Elsevier B.V. All rights reserved.

Study on core flow distribution of the reference core design Mark-III of experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Satoh, Sadao; Arai, Taketoshi; Miyamoto, Yoshiaki; Hirano, Mitsumasa

1977-01-01

Concerning the coolant flow distribution between fuel channels and other flow paths in the core, designated as Reference Core Mark-III of the Multi-purpose Experimental Very High Temperature Reactor, thermal analysis has been made of the control rods and other steel structures around the core to find the coolant flow rates (bypass flow) necessary to cool them to their safe operating temperatures. Calculations showed that adequate cooling could be achieved in the Mark-III Core by the bypass flow of 8% of the total reactor coolant flow, 4% each for the control-rod channels and for other structures. The thermal and coolant flow design bases, including the assumption of a 10% bypass flow, were thus confirmed to first approximation. (auth.)

Experimental fusion power reactor conceptual design study. Final report. Volume III

International Nuclear Information System (INIS)

Baker, C.C.

1976-12-01

This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

Innovative fission reactors for this century

International Nuclear Information System (INIS)

Minguez, E.

2007-01-01

It is well known that global trends indicate a rebirth of nuclear energy due to several items: the climate change and the use of energies that emits CO 2 , the cost and dependence of gas and oil, the new innovative reactors which are competitive, safer, and sustainable and can support the Kyoto Protocol. The Advanced Reactors have safer systems than those developed in the Generation II, which demonstrates that are sustainable for the present and nuclear industry has also developed new concepts for the future which also will be sustainable. Now the new power plants that have being constructed are classified in the Generation III. Several units of this technology are in operation in Japan and other countries of the Pacific. Europe is now constructing the first unit in Finland (Olkilouto) with European technology: the European Pressurized Reactor (EPR). France has announced the beginning of the construction of an EPR in Flamanville next year. In 2000, several countries with advanced nuclear technology established the Generation IV International Forum (GIF) to develop and demonstrate nuclear energy systems that offer advantages in the following areas: sustainability, economics, safety and reliability and proliferation resistance and physical protection. These new systems will be deployed commercially after 2030. Six innovative concepts are under research, and the aim is not only produce electricity, but also hydrogen using the operational conditions of several concepts. Developed countries with NPPs in operation have strategies for the future of the nuclear energy. For the short term is to extend the operation of the NPPs until 60 years, or alternatively construction of new units of Generation III, to substitute those closed for decommissioning, keeping the percentage of contribution to the electricity generated. Between the period 2030-50, the solution is to operate the new innovative systems of the Generation IV, which uses the passive concept, and in the second part

Safety analysis of RA Reactor operation I-III; Analiza sigurnosti rada Reaktora RA I - III, IZ-213-0322-1963

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

This safety analysis report covers the following three parts: Technical and operational characteristics of the RA reactor; Accidents analysis; and Environmental effects of the maximum possible accident. [Serbo-Croat] Ovaj izvestaj o analizi sigurnosti rada reaktora RA sastoji se od tri dela: Tehnicke i pogonske karakteristike reaktora RA; Analiza akcidenta; i Posledice maksimalno moguceg akcidenta na okolinu reaktora.

ITER at the international conference on fusion reactor materials

International Nuclear Information System (INIS)

Kalinin, G.; Barabash, V.; Matera, R.

1998-01-01

The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

HAHA--nothing to laugh about. Measuring the immunogenicity (human anti-human antibody response) induced by humanized monoclonal antibodies applying ELISA and SPR technology.

Science.gov (United States)

Nechansky, Andreas

2010-01-05

Immunogenicity induced by passively applied proteins is a serious issue because it is directly related to the patient's safety. The out-come of an immune reaction to a therapeutic protein can range from transient appearance of antibodies without any clinical significance to severe life threatening conditions. Within this article, enzyme-linked immunosorbent assay (ELISA) and surface plasmon resonance (SPR) methodology to measure immunogenicity are compared and the pros and cons are discussed.

Electrometallurgical treatment of aluminum-matrix fuels

International Nuclear Information System (INIS)

Willit, J.L.; Gay, E.C.; Miller, W.E.; McPheeters, C.C.; Laidler, J.J.

1996-01-01

The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% (volume basis) aluminum

Decontamination of fast reactor hulls and properties of immobilised waste forms

International Nuclear Information System (INIS)

Biddle, P.; Brown, P.E.; Rees, J.H.

1984-01-01

The objectives of this contract are threefold: (i) to characterise the activities associated with genuine hulls generated in reprocessing fuel from the Prototype Fast Reactor; (ii) to examine the feasibility of decontaminating the hulls by the use of simple low-temperature techniques; and (iii) to immobilise the hulls in cement and evaluate the products. Over the last year, work has concentrated in areas (i) and (ii), but the balance of the programme is now shifting more towards topic (iii). Progress is reported. (author)

Market share scenarios for Gen-DIII and gen-IV reactors in Europe

International Nuclear Information System (INIS)

Roelofs, F.; Heek, A. V.; Durpel, L. V. D.

2008-01-01

Nuclear energy is back on the agenda worldwide in order to meet growing energy demand and especially the growth in electricity demand. Many objectives direct to an increased use of nuclear energy, i.e. minimising energy costs, reducing climate change effects and others. In the light of the potential renewed growth of nuclear energy, the public demands a clear view on what nuclear energy may contribute towards meeting these objectives and especially how nuclear energy may address some socio-political obstructions with respect to economics, radioactive waste, safety and proliferation of fissile materials. To address these questions, the future nuclear reactor park mix in Europe has been analysed applying an integrated dynamic process modelling technique. Various market share scenarios for nuclear energy are derived including sub-variants with regard to the intra-nuclear options. In the analyses, it is assumed that different types of new reactors may be built, taking into account the introduction date of considered Gen-Ill (i.e. EPR) and Gen-IV (i.e. SCWR, HTR, FR) reactors, and the economic evaluation of the complete fuel cycle. The assessment was undertaken using the DANESS code (Dynamic Analysis of Nuclear Energy System Strategies). The analyses show that given the considered realistic nuclear energy demand and given a limited number of available Gen-III and Gen-IV reactor types, the future European nuclear park will exist of combinations of Gen-III and Gen-IV reactors. This mix will always consist of a set of reactor types each having its specific strengths. The analyses also highlight the triggers influencing the choice between different nuclear energy deployment scenarios. (authors)

A review of fast reactor program in Japan (April 2001 - March 2002)

International Nuclear Information System (INIS)

Nagata, T.; Ieda, Y.

2002-01-01

This report describes the research and development activities on fast reactors in Japan thru April 2001 to March 2002. In December 2001, the Cabinet decided the Plan for Reorganization of Government-funded Corporations including the merger of JNC and the Japan Atomic Energy Research Institute (JAERI). A law to set up a new entity is supposed to be submitted to the National Diet by the Japanese Fiscal Year (JFY) 2004. In the Experimental Fast Reactor Joyo, thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. Though a fire broke out in the maintenance building of Joyo in October 2001, the Mk-III construction work was restarted in February 2002. In the Prototype Fast Breeder Reactor Monju, countermeasures against sodium leakage have already been drawn up based on Monju comprehensive safety review. The safety licensing examination for the plant modification of Monju is undergoing. As for the Feasibility Study on Commercialized Fast Reactor Cycle Systems, JFY2001 was the first year of its second phase. A three-year period from JFY2001 to 2003 is the initial term of this phase. During this term, research activities are being focused on the design of the candidate concepts and fundamental tests of key technologies. An interim summary of these activities will be checked and reviewed, and based on the results; the research for JFY 2004 to 2005 will be conducted in order to narrow down the number of alternatives for the fast reactor cycle. (author)

Fe(III)-solar light induced degradation of diethyl phthalate (DEP) in aqueous solutions.

Science.gov (United States)

Mailhot, G; Sarakha, M; Lavedrine, B; Cáceres, J; Malato, S

2002-11-01

The degradation of diethyl phthalate (DEP) photoinduced by Fe(III) in aqueous solutions has been investigated under solar irradiation in the compound parabolic collector reactor at Plataforma Solar de Almeria. Hydroxyl radicals *OH, responsible of the degradation, are formed via an intramolecular photoredox process in the excited state of Fe(III) aquacomplexes. The primary step of the reaction is mainly due to the attack of *OH radicals on the aromatic ring. For prolonged irradiations DEP and its photoproducts are completely mineralized due to the regeneration of the absorbing species and the continuous formation of *OH radicals that confers a catalytic aspect to the process. Consequently, the degradation photoinduced by Fe(III) could be an efficient method of DEP removal from water.

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Science.gov (United States)

2010-01-01

... with Mark III type containments and all pressurized water reactors with ice condenser containments must... condenser containments that do not rely upon an inerted atmosphere inside containment to control combustible... containment atmosphere following a significant beyond design-basis accident for combustible gas control and...

Corrosion control in CANDU nuclear power reactors

International Nuclear Information System (INIS)

Lesurf, J.E.

1974-01-01

Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

Characteristics and economy of the European reactor of pressurized water (EPR); Caracteristicas y economia del reactor europeo de agua a presion (EPR)

Energy Technology Data Exchange (ETDEWEB)

Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jov@nuclear.inin.mx

2005-07-01

The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

RA Research reactor, Part I: Technical and operational properties of the RA reactor; Analiza sigurnosti rada Reaktora RA I-III, Deo I: Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Zecevic, V; Nikolic, M; Popovic, B; Milosevic, M; Milic, M; Strugar, P; Pesic, M; Nikolic, V; Rajic, M; Radivojevic, J; Jankovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

RA reactor is a research reactor with rather high power density. Apart from research it is used for isotope production and industrial applications due to high reactivity excess (about 11%). It is a thermal reactor, heavy water moderated, cooled by D{sub 2}O, and H{sub 2}O, with a graphite reflector. Nominal power is 6.5 MW. Fuel is 2% enriched metal uranium, reactor core height is 1220 mm, and diameter is 1405 mm. Reactor lattice is square with lattice pitch 130 mm. There is 6 horizontal experimental channels and a graphite column. There is a total of 84 fuel channels and 45 experimental channels in the core. Maximum thermal neutron flux is 5.5 10{sup 13} n/cm{sup 2} s at nominal power level.

Accident progression event tree analysis for postulated severe accidents at N Reactor

International Nuclear Information System (INIS)

Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M.; Medford, G.T.

1990-06-01

A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied

RA reactor safety analysis, Part II - Accident analysis; Analiza sigurnosti rada Reaktora RA I-III, Deo II - Analiza akcidenta

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Radanovic, Lj; Milovanovic, M; Afgan, N; Kulundzic, P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

This part of the RA reactor safety analysis includes analysis of possible accidents caused by failures of the reactor devices and errors during reactor operation. Two types of accidents are analyzed: accidents resulting from uncontrolled reactivity increase, and accidents caused by interruption of cooling.

Analysis of a Spanish energy scenario with Generation IV nuclear reactors

International Nuclear Information System (INIS)

Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

2013-01-01

Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

Hydrodynamic cavitation to improve bulk fluid to surface mass transfer in a nonimmersed ultraviolet system for minimal processing of opaque and transparent fluid foods.

Science.gov (United States)

Milly, P J; Toledo, R T; Chen, J; Kazem, B

2007-11-01

Ultraviolet (UV)-induced chemical reactions and inactivation of microorganisms in transparent and opaque fluids are strongly dependent upon the homogenous exposure of the target species to the UV irradiation. Current UV technologies used in water disinfection and food preservation applications have limited efficacy due to suspended particles shading target species. An Ultraviolet-Shockwave Power Reactor (UV-SPR) consisting of an inner rotating rotor and a stationary quartz housing and 2 end plates was used to induce 'controlled cavitation.' Eight UV low-pressure mercury lamps spaced uniformly were installed lengthwise around the quartz housing periphery. A KI to I(3) (-)chemical dosimeter for UV was used to quantify photons received by fluid in the annular space of the SPR. UV dose (J/m(2)) increased from 97 J/m(2) at 0 rpm to over 700 J/m(2) for SPR speeds above 2400 rpm. Inactivation of E. coli 25922 in apple juice and skim milk in the UV-SPR at exit temperatures below 45 degrees C was greater than 4.5 and 3 logs, respectively. The UV-SPR system proved successful in increasing the mass transfer of transparent and opaque fluid to the UV irradiated surface.

The Jules Horowitz Reactor project, a driver for revival of the research reactor community

International Nuclear Information System (INIS)

Pere, P.; Cavailler, C.; Pascal, C.

2010-01-01

The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first-rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10 14 n.cm -2 /sec -1 E>1 MeV in core and 3,6x10 14 n.cm -2 /sec -1 E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960s, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items:accommodation of a broad experimental domain; involvement by international partners making in-kind contributions to the project; ? development of components critical to safety and performance; the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and; tools to carry out the project, including computer codes

The determination of the space distribution, energy spectrum and dose parameters of thermal column beam resulting from swimming pool reactor

International Nuclear Information System (INIS)

Chen Changmao; Xie Jianlun; Leng Ruiping; Song Shushou; Su Jingling

1991-01-01

The axial and radial distribution, epithermal energy spectrum and dose equivalent rate of thermal column beam resulting from SPR have been determined in the Institute of Atomic Energy. The results show that the neutron fluence rate along the axial direction decreases as the distance increases outside the thermal column channel, and the trend of fluence rate attenuation follows approximately the inverse square law of a point source. When the reactor thermal power rate is 3 MW, at a distance of 50 cm to the channel, the thermal and epithermal neutron fluence rate are about 1.61 x 10 7 and 6.1 x 10 4 n/cm 2 · s respectively; dose equivalent rates are some 62 and 2.9 cSv/h respectively. At the end of the chennal, γ dose equivalent rate is 60 cSv/h or so

Decommissioning of TRIGA Mark II type reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

Liquid plugs bouncing against a solid basis, comparison of SIMMER-III and PLEXUS results; Zum Aufprall von Fluessigkeitssaeulen auf starre Hindernisse Vergleich von SIMMER-III und PLEXUS Ergebnissen

Energy Technology Data Exchange (ETDEWEB)

Maschek, W.; Arnecke, G.; Flad, M.

1995-08-01

The SIMMER III code is being tested by application to problems of reactor accidents. The fluid dynamics/thermohydraulics part of the code can be applied also to evaporation/condensation, melt, and freezing phenomena. For a liquid plug bouncing against a solid basis, the momentum transfer is calculated. PLEXUS results turn out to be in significant disagreement with the SIMMER calculations. (orig.)

Latest Results from the Daya Bay Reactor Neutrino Experiment

CERN Multimedia

CERN. Geneva

2014-01-01

Among all the fundamental particles that have been experimentally observed, neutrinos remain one of the least understood. The Daya Bay Reactor Neutrino Experiment in China consists of eight identical detectors placed underground at different baselines from three groups of nuclear reactors, a configuration that is ideally suited for studying the properties of these elusive particles. This talk will present three sets of results that have just recently been released by the Daya Bay Collaboration: (i) a precision measurement of the oscillation parameters that drive the disappearance of electron antineutrinos at short baselines, (ii) a search for sterile neutrino mixing, and (iii) a high-statistics determination of the absolute flux and spectrum of reactor-produced electron antineutrinos. All of these results extend the limits of our knowledge in their respective areas and thus shed new light on neutrinos and the physics that surround them.

Analysis and application of a simulator of a nuclear reactor AP-600; Analisis y aplicacion de un simulador de un reactor nuclear AP-600

Energy Technology Data Exchange (ETDEWEB)

Medina S, V. S. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: medina_victor@comunidad.unam.mx [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (MX)

2011-11-15

In front of the resurgence of interest in the nuclear power production, several national organizations have considered convenient to have highly specialized human resources in the technologies of nuclear reactors of III + and IV generation. For this task, the intensive and extensive applications of the computation should been considered, as the virtual instrumentation. The present work analyzes the possible applications of a nuclear simulator provided by the IAEA with base in the design of the reactor AP-600, using a focusing of modular model developed in FORTRAN. One part of the work that was made with the simulator includes the evaluation of 21 transitory events of operation, including the recreation of the accident happened in the nuclear power plant of Three Mile Island in 1979, comparing the actions flow and the answer of the systems under the intrinsic security of a III + generation reactor. The impact that had the mentioned accident was analyzed in the growing of the nuclear energy sector and in the public image with regard to the nuclear power plants. An application for this simulator was proposed, its use as tool for the instruction in the nuclear engineering courses using it to observe the operation of the different security systems and its interrelation inside the power plant as well as a theoretical/practical approach for the student. (Author)

ATWS: a reappraisal. Part III. Frequency of anticipated transients. Interim report

International Nuclear Information System (INIS)

Leverenz, F.L. Jr.; Koren, J.M.; Erdmann, R.C.; Lellouche, G.S.

1978-07-01

The document is Part III of the Institute study of the ATWS question. The frequencies of the various events which have led to a reactor scram are documented from the nuclear power plant records. Some of these events, in the absence of scram, could lead to undesirable system response and are the ''transients of significance'' which comprise the anticipated transients of the ATWS question

A user input manual for single fuel rod behaviour analysis code FEMAXI-III

International Nuclear Information System (INIS)

Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

1983-03-01

Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

International Nuclear Information System (INIS)

Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

1982-01-01

Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

Investigation reactor D-2201 polypropylene production unit using nuclear technique

International Nuclear Information System (INIS)

Wibisono; Sugiharto; Jefri Simanjuntak

2016-01-01

D-2201 reactor is a unit in the polypropylene production process at Pertamina Refinery Unit III Plaju. Reactor with a capacity of 45 kilo liter is not operated in normal operation condition. The validity of liquid level indicator on the unit is doubtful when refers to the production quality. Gamma source of 150 mCi Cobalt-60 and a scintillation detector had been used to scan the outer wall of the reactor to detect the liquid level during operation with a capacity of 40 %. Measurements were made along the reactor walls with 25 mm scan resolution and 5 seconds time sampling. Experiment result shows that the liquid level at the position of 40 % and at normal level position are not observed. Investigation did not find the liquid level above normal. D-2201 is diagnose not normal operating condition diagnosed with liquid abundant passed the recommended limits. Investigation advised to repair or to calibrate the liquid level indicator which is currently installed. (author)

VTT ENIGMA Calculations for FUMEX-III CRP

International Nuclear Information System (INIS)

Tulkki, Ville

2013-01-01

International Atomic Energy Agency IAEA has initiated a string of Coordinated Research Programmes (CRPs) to enhance co-operation between fuel modellers. One of these CRPs, FUMEX-III, was ongoing during 2008 - 2011 and has provided material and incentive for assessment of the fuel codes. This report presents the Finnish FUMEX-III simulations performed with the VTT-modified ENIGMA v5.9b. The work has been done as a part of SAFIR2010 (SAfety of FInnish Reactors 2010) project POKEVA (years 2008 to 2010) and SAFIR2014 project PALAMA (year 2011). VTT Technical Research Centre of Finland received ENIGMA v5.9b from Nuclear Electric plc of the UK in 1992. Internal development has been on-going since then. 'ENIGMA v5.9b with VTT modifications' (from now on, 'ENIGMA' for short) is a separate and different program from British Energy's ENIGMA 5.14 and UK National Nuclear Laboratory's ENIGMA-B. The FUMEX-III work has been performed in tandem of VTT's internal review work attempting to catalogue the changes done since 1992 and to assess the current state of the code. Several individual internal reports detail the changes made and the individual model assessments done during the years. (author)

Development of an SPR imaging biosensor for determination of cathepsin G in saliva and white blood cells

International Nuclear Information System (INIS)

Gorodkiewicz, E.; Wojtulewski, K.; Regulska, E.

2011-01-01

Cathepsin G (CatG) is an endopeptidase that is associated with the early immune response. The synthetic compound cathepsin G inhibitor I (CGI-I) was tested for its ability to inhibit the activity of CatG via a new surface plasmon resonance imaging assay. CGI-I was immobilized on the gold surface of an SPR sensor that was first modified with 1-octadecanethiol. A concentration of CGI-I equal to 4.0 μg.mL -1 and a pH of 8.0 were found to give the best results. The dynamic response of the sensor ranges from 0. 25 to 1. 5 ng.mL -1 , and the detection limit is 0. 12 ng.mL -1 . The sensor was applied to detect CatG in human saliva and white blood cells. (author)

Self-assessment of application of the Code of Conduct on the safety of research reactors - Mexico

International Nuclear Information System (INIS)

Mamani-Alegria, Y.R.; Salgado-Gonzalez, J.R.; Miranda-Aldaco, J.

2009-01-01

In Mexico, the nuclear regulatory body is the National Commission on Nuclear Safety and Safeguards (CNSNS), and there is one research reactor, a TRIGA MARK III, operated by the National Institute for Nuclear Research (ININ). The main aspects of the Self-assessment of application of The Code of Conduct on the Safety of Research Reactor are given for the case of the TRIGA reactor. Furthermore, in this paper we give a brief description of the legal framework of the licensing process, for nuclear activities in a research reactor, there are also highlights of the major reactor features, the uses of the reactor for isotope production, the management and verification of safety, the radiation protection management program, the emergency planning and the training and qualification of the operation personnel. (author)

Burnable poison calculations for Mk.III gas-cooled reactor systems

Energy Technology Data Exchange (ETDEWEB)

Gubbins, M E

1971-02-15

A method of calculating the reactivity and burn-up hisotry of a Mk.III GCR system containing burnable poisons has been described. The method allows for poison-fuel interaction. Using the method it has been shown that burn-up of the poison under a constant incident flux can give errors of the order of 1-2 niles. A calculation using the method described will take about 50% longer than a straightforward fuel burn-up calculation in the same number of groups. The multi-cell approach has a potential for handling greater geometrical complexity. It is intended to compare the method against experiment as soon as suitable experimental results become available.

Isothiocyanato complexes of Gd(III), Tb(III), Dy(III) and Ho(III) with 2-(2'-pyridyl)benzimidazole

Energy Technology Data Exchange (ETDEWEB)

Mishra, A; Singh, V K

1982-01-01

Six-coordinated complexes of the type (Ln(PyBzH)/sub 2/NCS.H/sub 2/O) (NCS)/sub 2/.nH/sub 2/O/mC/sub 2/H/sub 5/OH (Ln = Gd(III), Tb(III), Dy(III) and Ho(III), n=1-2; m=1) have been prepared from Ln(NCS)/sub 6//sup 3 -/. The room temperature magnetic moment values confirm the terpositive state of the lanthanide ions. Infrared spectra suggest the N-coordination of thiocyanate group. Electronic spectral studies of Tb(III), Dy(III) and Ho(III) complexes have been made in terms of LSJ term energies. 13 refs.

Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2007-01-01

The Global Nuclear Energy Partnership (GNEP) seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy. Fully meeting the GNEP vision may require the deployment of thousands of reactors in scores of countries, many of which do not use nuclear energy currently. Some of these needs will be met by large-scale Generation III and III+ reactors (>1000 MWe) and Generation IV reactors when they are available. However, because many developing countries have small and immature electricity grids, the currently available Generation III(+) reactors may be unsuitable since they are too large, too expensive, and too complex. Therefore, GNEP envisions new types of reactors that must be developed for international deployment that are 'right sized' for the developing countries and that are based on technologies, designs, and policies focused on reducing proliferation risk. The first step in developing such systems is the generation of technical requirements that will ensure that the systems meet both the GNEP policy goals and the power needs of the recipient countries. Reactor systems deployed internationally within the GNEP context must meet a number of requirements similar to the safety, reliability, economics, and proliferation goals established for the DOE Generation IV program. Because of the emphasis on deployment to nonnuclear developing countries, the requirements will be weighted differently than with Generation IV, especially regarding safety and non-proliferation goals. Also, the reactors should be sized for market conditions in developing countries where energy demand per capita, institutional maturity and industrial infrastructure vary considerably, and must utilize fuel that is compatible with the fuel recycle technologies being developed by GNEP. Arrangements are already underway to establish Working Groups jointly with Japan and Russia to develop requirements for reactor systems. Additional bilateral and multilateral

An advanced conceptual Tokamak fusion power reactor utilizing closed cycle helium gas turbines

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

UWMAK-III is a conceptual Tokamak reactor designed to study the potential and the problems associated with an advanced version of Tokamaks as power reactors. Design choices have been made which represent reasonable extrapolations of present technology. The major features are the noncircular plasma cross section, the use of TZM, a molybdenum based alloy, as the primary structural material, and the incorporation of a closed-cycle helium gas turbine power conversion system. A conceptual design of the turbomachinery is given together with a preliminary heat exchanger analysis that results in relatively compact designs for the generator, precooler, and intercooler. This paper contains a general description of the UWMAK-III system and a discussion of those aspects of the reactor, such as the burn cycle, the blanket design and the heat transfer analysis, which are required to form the basis for discussing the power conversion system. The authors concentrate on the power conversion system and include a parametric performance analysis, an interface and trade-off study and a description of the reference conceptual design of the closed-cycle helium gas turbine power conversion system. (Auth.)

Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5; Calculo de activacion neutronica de barras de control de un reactor nuclear, utilizando MCNP5

Energy Technology Data Exchange (ETDEWEB)

Pena V, J.D.

2016-07-01

The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

Transients in reactors for power systems compensation

Science.gov (United States)

Abdul Hamid, Haziah

surge arrester operation during the MSCDN energisation, which causes steep voltage change at the reactor terminal. (ii) Second, the nonuniform voltage distribution, resulting in high stresses across the top inter-turn windings. (iii) Third, the rapid rate-of-change of voltage in the assumed worst-case reactor winding location. This is accompanied by a high dielectric current through the inter-turn winding insulation..

Complexation of trivalent actinides and lanthanides with hydrophilic N-donor ligands for Am(III)/Cm(III) and An(III)/Ln(III) separation; Komplexierung von trivalenten Actiniden und Lanthaniden mit hydrophilen N-Donorliganden zur Am(III)/Cm(III)- bzw. An(III)/Ln(III)-Trennung

Energy Technology Data Exchange (ETDEWEB)

Wagner, Christoph

2017-07-24

The implementation of actinide recycling processes is considered in several countries, aiming at the reduction of long-term radiotoxicity and heat load of used nuclear fuel. This requires the separation of the actinides from the fission and corrosion products. The separation of the trivalent actinides (An(III)) Am(III) and Cm(III), however, is complicated by the presence of the chemically similar fission lanthanides (Ln(III)). Hydrophilic N-donor ligands are employed as An(III) or Am(III) selective complexing agents in solvent extraction to strip An(III) or Am(III) from an organic phase loaded with An(III) and Ln(III). Though they exhibit excellent selectivity, the complexation chemistry of these ligands and the complexes formed during solvent extraction are not sufficiently characterized. In the present thesis the complexation of An(III) and Ln(III) with hydrophilic N-donor ligands is studied by time resolved laser fluorescence spectroscopy (TRLFS), UV/Vis, vibronic sideband spectroscopy and solvent extraction. TRLFS studies on the complexation of Cm(III) and Eu(III) with the Am(III) selective complexing agent SO{sub 3}-Ph-BTBP (tetrasodium 3,3{sup '},3'',3{sup '''}-([2,2{sup '}-bipyridine]-6,6{sup '}-diylbis(1,2,4-triazine-3,5,6-triyl)) tetrabenzenesulfonate) revealed the formation of [M(SO{sub 3}-Ph-BTBP){sub n}]{sup (4n-3)-} complexes (M = Cm(III), Eu(III); n = 1, 2). The conditional stability constants were determined in different media yielding two orders of magnitude larger β{sub 2}-values for the Cm(III) complexes, independently from the applied medium. A strong impact of ionic strength on the stability and stoichiometry of the formed complexes was identified, resulting from the stabilization of the pentaanionic [M(SO{sub 3}-Ph-BTBP){sub 2}]{sup 5-} complex with increasing ionic strength. Thermodynamic studies of Cm(III)-SO{sub 3}-Ph-BTBP complexation showed that the proton concentration of the applied medium impacts

The utility industry and reactor surveillance

International Nuclear Information System (INIS)

Jenkins, R.B.

1983-01-01

Every commercial nuclear power reactor pressure vessel (RPV) is required to have a reactor vessel surveillance program at the time of plant licensing. The program is part of a continuing structural integrity assessment of the RPV. As such, the surveillance program supplements Section III of the American Society of Mechanical Engineers (ASME) Code (1), which is the design basis for nuclear power plant component pressure boundaries. The Code assumes that the materials of construction are ductile in the evaluation and design of all components. The surveillance program for each RPV is intended to provide assurance of continued applicability of the ASME Code, Appendix G, assessment of that RPV's operating limits. This assessment ensures that the RPV is always in a condition which precludes the unstable propagation of flaws in the vessel wall material. The potential presence of flaws and the desire to ensure ductility are significant considerations in ferritic steels such as those used to fabricate nuclear reactor pressure vessels. These materials are known to exhibit transition from ductile-to-brittle fracture behavior over a determined temperature range. Neutron irradiation tends to shift this ductile-to-brittle behavior transition zone to a temperature higher than unirradiated materials

Generic Procedures for Response to a Nuclear or Radiological Emergency at Triga Research Reactors. Attachment 1 (2011)

International Nuclear Information System (INIS)

2011-01-01

The publication provides guidance for response to emergencies at TRIGA research reactors in Threat Category II and III. It contains information on the unique behaviour of TRIGA fuel during accident conditions; it describes design characteristics of TRIGA research reactors and provides specific symptom-based emergency classification for this type of research reactor. This publication covers the determination of the appropriate emergency class and protective actions for a nuclear or radiological emergency at TRIGA research reactors. It does not cover nuclear security at TRIGA research reactors. The term 'threat category' is used in this publication as described in Ref. [6] and for the purposes of emergency preparedness and response only; this usage does not imply that any threat, in the sense of an intention and capability to cause harm, has been made in relation to facilities, activities or sources. The threat category is determined by an analysis of potential nuclear and radiological emergencies and the associated radiation hazard that could arise as a consequence of those emergencies. STRUCTURE. The attachment consists of an introduction which defines the background, objective, scope and structure, two sections covering technical aspects and appendices. Section 2 describes the characteristics of TRIGA fuel in normal and accident conditions. Section 3 contains TRIGA research reactor specific emergency classification tables for Threat Category II and III. These tables should be used instead of the corresponding emergency classification tables presented in Ref. [1] while developing the emergency response arrangements at TRIGA research reactors. The appendices present some historical overview and typical general data for TRIGA research reactor projects and the list of TRIGA installations around the world. The terms used in this document are defined in the IAEA Safety Glossary and the IAEA Code of Conduct on the Safety of Research Reactors.

Safety of light water reactors. Risks of nuclear technology

International Nuclear Information System (INIS)

Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

2012-01-01

The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor; Desarrollo de un simulador para diseno y prueba de controladores de potencia en un reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Perez M, C.; Benitez R, J.S.; Lopez C, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

Complexes of 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III)

International Nuclear Information System (INIS)

Ferenc, W.; Bernat, M; Gluchowska, H.W.; Sarzynski, J.

2010-01-01

The complexes of 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III) have been synthesized as polycrystalline hydrated solids, and characterized by elemental analysis, spectroscopy, magnetic studies and also by X-ray diffraction and thermogravimetric measurements. The analysed complexes have the following colours: violet for Nd(III), white for Gd(III) and cream for Ho(III) compounds. The carboxylate groups bind as bidentate chelating (Ho) or bridging ligands (Nd, Gd). On heating to 1173K in air the complexes decompose in several steps. At first, they dehydrate in one step to form anhydrous salts, that next decompose to the oxides of respective metals. The gaseous products of their thermal decomposition in nitrogen were also determined and the magnetic susceptibilities were measured over the temperature range of 76-303K and the magnetic moments were calculated. The results show that 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III) are high-spin complexes with weak ligand fields. The solubility value in water at 293K for analysed 4-chlorophenoxyacetates is in the order of 10 -4 mol/dm 3 . (author)

Hazards classification determination for PUREX fuel transfer to K-Basins

International Nuclear Information System (INIS)

Dodd, E.N. III.

1995-01-01

The PUREX Plant presently contains 2.9 metric tons of an aluminum clad Single Pass Reactor (SPR) fuel which is stored under water in four open top buckets in the PUREX slug storage basin. The PUREX dissolver cells contain approximately 0.5 metric tons of zirconium clad N Reactor fuel which was inadvertently placed into the process cell during charging operations. The dissolver N reactor elements will be recovered from the process floors using new crane operated tools. When the fuel shipment(s) is scheduled, the cask cars will be positioned into the PUREX rail tunnel and the overhead door will be opened. All the SPR fuel will be loaded into two cask rail cars inside four casks. The N Reactor fuel will be loaded into a separate rail car inside two or three casks. The car loading is initiated by opening the rail car lid and removing the cask lids. Prior to loading the canisters of N Reactor fuel, the canisters will be refilled with water (as needed) and a lid will be installed. The baskets of SPR fuel or canisters of N Reactor fuel will then be loaded into the casks. The lids to the casks will then be reinstalled and the car lids closed. The rail cars will then be decontaminated as necessary. The cask cars will be shipped either in two shipments or a combined single shipment using the rail route between PUREX and the K Basins. At the basin, the cask car will be positioned in the loadout area. The cask car lid will be opened and a single cask moved into the loadout pit, which is a lowered section of the basin. The cask lid is removed while the cask is lower into the pit. The fuel is then removed from the cask and stored in the basin. The cask is then removed, the lid reinstalled during removal, and the cask replaced into the cask car. This document identifies the hazard classification of the Fuel Transfer from the PUREX facility to K-Basins

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)

Decontamination and decommissioning project status of the TRIGA mark-2±3 research reactors

International Nuclear Information System (INIS)

Jung, K. J.; Baek, S. T.; Jung, W. S.; Park, S. K.; Jung, K. H.

1999-01-01

TRIGA Mark-II, the first research reactor in Korea, has operated since 1962, and the second one, TRIGA Mark-III since 1972. Both of them had their operation phased out in 1995 due to their lives and operation of the new research reactor, HANARO at the Korea Atomic Energy Research Institute (KAERI) in Taejeon. Decontamination and decommissioning (D and D) project of the TRIGA Mark-II and Mark-III was started in January 1997 and will be completed in December 2002. In the first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of Korea Institute of Nuclear Safety (KINS). In 1998, Hyundai Engineering Company (HEC) is the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels plc (BNFL) is technical assisting partner of HEC. The decommissioning plan document was submitted to the Ministry of Science and Technology (MOST) for the decommissioning license in December 1998, and it expecting to be issued a license at the end of September 1999. The goal of this project is to release the reactor site and buildings as an unrestricted area. This paper summarizes current status and future plan for the D and D project

Site-directed antibody immobilization using a protein A-gold binding domain fusion protein for enhanced SPR immunosensing.

Science.gov (United States)

de Juan-Franco, Elena; Caruz, Antonio; Pedrajas, J R; Lechuga, Laura M

2013-04-07

We have implemented a novel strategy for the oriented immobilization of antibodies onto a gold surface based on the use of a fusion protein, the protein A-gold binding domain (PAG). PAG consists of a gold binding peptide (GBP) coupled to the immunoglobulin-binding domains of staphylococcal protein A. This fusion protein provides an easy and fast oriented immobilization of antibodies preserving its native structure, while leaving the antigen binding sites (Fab) freely exposed. Using this immobilization strategy, we have demonstrated the performance of the immunosensing of the human Growth Hormone by SPR. A limit of detection of 90 ng mL(-1) was obtained with an inter-chip variability lower than 7%. The comparison of this method with other strategies for the direct immobilization of antibodies over gold surfaces has showed the enhanced sensitivity provided by the PAG approach.

Present and future use of TRIGA reactors in the Federal Republic of Germany

International Nuclear Information System (INIS)

Menke, H.; Junker, D.; Krauss, O.

1986-01-01

In the Federal Republic of Germany nine research reactors are presently in operation, three of which are TRIGA reactors. These are the TRIGA Mark I reactors at Hannover and Heidelberg with a steady state power of 250 kW and the TRIGA Mark II reactor at Mainz with a steady power of 100 kW and a peak pulsing power of 250 MW. The decommissioning of a number of research reactors, including the TRIGA Mark III reactor at Neuherberg near Munich, is reason enough to think about the present and future use of our reactors. The German TRIGA reactors met a lively interest of scientists, since they went into operation. Presently they are well used especially in biomedical (Hannover, Heidelberg) and basic research (Mainz). In the course of about 20 years of operation the techniques and requirements of experiments changed and consequently the use of the reactors too. Certainly this will be so in the future. But thanks to its versatile experimental facilities, this type of reactor can meet the various experimental demands. So we are looking forward to a good utilisation of our German TRIGA reactors in future and taking into account the low costs for personal, energy and fuel, we are quite confident that they will be in operation still for many years. (author)

The Jules Horowitz reactor project, a driver for revival of the research reactor community

Energy Technology Data Exchange (ETDEWEB)

Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

2010-07-01

The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

An assessment of methods of calculating sodium-voiding reactivity in plutonium-fuelled fast reactors

International Nuclear Information System (INIS)

Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

1980-01-01

After a survey of the requirements an assessment of the accuracy of calculations of the sodium-void effect using UK methods and data is made on the basis of the following work: (a) The analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(e)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first-order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. (b) Theoretical studies of some effects, including the following: (i) The effects of extrapolating to fuel operating temperature; (ii) Fuel-cycle and burnup effects, including the gradual replacement through a fuel cycle of control-rod absorption by fission product absorption, the loss of fissile material and the change in fuel nuclide relative composition; (iii) The heterogeneity effects of large fuelled subassemblies in pin geometry. (c) Theoretical studies of approximations in the calculational methods, including the following: (i) The importance in the whole reactor calculation of the energy group structure and the spatial mesh, including comparisons of calculations in two (RZ) and three-dimensional geometry; (ii) The importance of reactor material boundaries in the calculation of resonance shielding effects; (iii) The use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (author)

Radiological protection of the staff during the decommissioning operations of the Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Ene, D.C.

2002-01-01

Dose rate estimates for periods of 100 days and 6, 10, 25, 100 years after the shut down of the Romanian VVR-S reactor are presented in this paper for some foreseen decommissioning activities which include: i) cutting the water pipe in the pump room and the reactor sealing operations; ii) extracting reactor components; and iii) handling and dismantling the internal structures taken of from the reactor. For the reactor components extracted from the reactor, the considered calculation points were placed in the central plan of the items, on the surface and at distances from the surface which correspond to +0.2m, +1m, +2m, +8m, and +10m. Time dependence of the resulted dose rates are presented and discussed. Qualitative comparison with the measured values from other VVR-S reactors is done. The obtained results assist to develop working procedures that must be observed during the decommissioning activities. (author)

Applicability of Coupled Thermalhydraulic Codes for Safety Analysis of Nuclear Reactors

International Nuclear Information System (INIS)

Gairola, A.; Bhowmik, P. K.; Shamim, J. A.; Suh, K. Y.

2014-01-01

To this end computational codes like RELAP and TRACE are used to model thermal-hydraulic response of nuclear power plant during an accident. By careful modeling and significant user experience these system codes are able to simulate the behavior of primary system and the containment to a reasonable extent. Comparatively decoupled simulation is simple but might not produce reality and the physics involved in an accurate manner. Thus simulation using two different system codes is interesting as the whole system is coupled through the pressure in the containment and flow through the break. Using this methodology it might be possible to get new insight about the primary and containment behavior by the precise simulation of the accident both in the current reactors and future Gen-III/III+ reactors. Couple thermalhydraulic code methodology is still new and require further investigations. Applicability of such methodology to the GEN-II plants have met with limited success, however a number of situations in which this methodology could be applied are still unexplored and thus provides a room for improvement and modifications

Applicability of Coupled Thermalhydraulic Codes for Safety Analysis of Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

Gairola, A.; Bhowmik, P. K.; Shamim, J. A.; Suh, K. Y. [Seoul National Univ., Seoul (Korea, Republic of)

2014-10-15

To this end computational codes like RELAP and TRACE are used to model thermal-hydraulic response of nuclear power plant during an accident. By careful modeling and significant user experience these system codes are able to simulate the behavior of primary system and the containment to a reasonable extent. Comparatively decoupled simulation is simple but might not produce reality and the physics involved in an accurate manner. Thus simulation using two different system codes is interesting as the whole system is coupled through the pressure in the containment and flow through the break. Using this methodology it might be possible to get new insight about the primary and containment behavior by the precise simulation of the accident both in the current reactors and future Gen-III/III+ reactors. Couple thermalhydraulic code methodology is still new and require further investigations. Applicability of such methodology to the GEN-II plants have met with limited success, however a number of situations in which this methodology could be applied are still unexplored and thus provides a room for improvement and modifications.

ABWR1. A Generation III.7 reactor after the Fukushima Daiichi accident

International Nuclear Information System (INIS)

Sato, Takashi; Matsumoto, Keiji; Kurosaki, Toshikazu; Taguchi, Keisuke

2015-01-01

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. The NSSS is exactly the same as that of the current ABWR. As for safety design it has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7 day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the CV building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most

Radioactive waste management at WWER type reactors

International Nuclear Information System (INIS)

1993-05-01

This report was prepared within the framework of the Technical Assistance Regional Project on Advice on Waste Management at WWER Type Reactors, which was initiated by the IAEA in 1991. The Regional Project is an integral part of the IAEA's activities directed towards improvement of the safety and reliability of nuclear power plants with WWER type reactors (Soviet designed PWRs). Forty-five WWER type units are currently in operation and twenty-five are under construction in Bulgaria, Czechoslovakia, Finland, Hungary and the former USSR. The idea of regional collaboration between eastern European countries under the auspices of the IAEA was discussed for the first time during the last meeting of the Council for Mutual Economic Assistance (CMEA) on spent fuel and radioactive waste management, held in Rez, Czechoslovakia, in October 1990. Since then, the CMEA and some of its former Member States have ceased to exist. However, there are many reasons for eastern European countries to continue their regional collaboration at a higher level. The USSR, the designer and supplier of WWER type reactors in eastern European countries, participated in the first phase of the project. The majority of WWER type reactors are situated in States of the former USSR (Russia and Ukraine). The main results of the first phase of the Regional Project are: (i) Re-establishment of communication channels among eastern European countries operating WWER type reactors by incorporating the IAEA's technical assistance; (ii) Identification of common waste management problems (administrative and technical) requiring resolution; (iii) Familiarization with radioactive waste management systems at nuclear power plants with WWER type reactors - Paks (Hungary), Loviisa (Finland), Jaslovske Bohunice (Czechoslovakia) and Novovoronezh (Russian Federation). Tabs

Formation constants of Sm(III), Dy(III), Gd(III), Pr(III) and Nd(III) complexes of tridentate schiff base, 2-[(1H-benzimidazol-2-yl-methylene) amino] phenol

International Nuclear Information System (INIS)

Omprakash, K.L.; Chandra Pal, A.V.; Reddy, M.L.N.

1982-01-01

A new tridentate schiff base, 2- (1H-benzimidazol-2-yl-methylene)amino phenol derived from benzimididazole-2-carbo-xaldehyde and 2-aminophenol has been synthesised and characterised by spectral and analytical data. Proton-ligand formation constants of the schiff base and metal-ligand formation constants of its complexes with Sm(III), Dy(III), Gd(III), Nd(III) and Pr(III) have been determined potentiometrically in 50% (v/v) aqueous dioxane at an ionic strength of 0.1M (NaClO 4 ) and at 25deg C using the Irving-Rossotti titration technique. The order of stability constants (logβ 2 ) is found to be Sm(III)>Dy(III)>Gd(III)>Pr(III)>Nd(III). (author)

Formation constants of Sm(III), Dy(III), Gd(III), Pr(III) and Nd(III) complexes of tridentate schiff base, 2-((1H-benzimidazol-2-yl-methylene) amino) phenol

Energy Technology Data Exchange (ETDEWEB)

Omprakash, K L; Chandra Pal, A V; Reddy, M L.N. [Osmania Univ., Hyderabad (India). Dept. of Chemistry

1982-03-01

A new tridentate schiff base, 2- (1H-benzimidazol-2-yl-methylene)amino phenol derived from benzimididazole-2-carbo-xaldehyde and 2-aminophenol has been synthesised and characterised by spectral and analytical data. Proton-ligand formation constants of the schiff base and metal-ligand formation constants of its complexes with Sm(III), Dy(III), Gd(III), Nd(III) and Pr(III) have been determined potentiometrically in 50% (v/v) aqueous dioxane at an ionic strength of 0.1M (NaClO/sub 4/) and at 25deg C using the Irving-Rossotti titration technique. The order of stability constants (log..beta../sub 2/) is found to be Sm(III)>Dy(III)>Gd(III)>Pr(III)>Nd(III).

Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

International Nuclear Information System (INIS)

Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

1996-01-01

Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

Conceptual design of a commercial accelerator driven thorium reactor

International Nuclear Information System (INIS)

Fuller, C. G.; Ashworth, R. W.

2010-01-01

This paper describes the substantial work done in underpinning and developing the concept design for a commercial 600 MWe, accelerator driven, thorium fuelled, lead cooled, power producing, fast reactor. The Accelerator Driven Thorium Reactor (ADTR TM) has been derived from original work by Carlo Rubbia. Over the period 2007 to 2009 Aker Solutions commissioned this concept design work and, in close collaboration with Rubbia, developed the physics, engineering and business model. Much has been published about the Energy Amplifier concept and accelerator driven systems. This paper concentrates on the unique physics developed during the concept study of the ADTR TM power station and the progress made in engineering and design of the system. Particular attention is paid to where the concept design has moved significantly beyond published material. Description of challenges presented for the engineering and safety of a commercial system and how they will be addressed is included. This covers the defining system parameters, accelerator sizing, core and fuel design issues and, perhaps most importantly, reactivity control. The paper concludes that the work undertaken supports the technical viability of the ADTR TM power station. Several unique features of the reactor mean that it can be deployed in countries with aspirations to gain benefit from nuclear power and, at 600 MWe, it fits a size gap for less mature grid systems. It can provide a useful complement to Generation III, III+ and IV systems through its ability to consume actinides whilst at the same time providing useful power. (authors)

Revisiting homogeneous suspension reactors for production of radioisotopes

International Nuclear Information System (INIS)

Pasqualini, E.E.

2010-01-01

Some 50 years ago in Geneva Conferences I, II and III (1955. 1958 and 1964) on the Peaceful Uses of Atomic Energy, and also in Vienna Symposium on Reactor Experiments (1961), several papers where presented by different countries referring to advances in homogeneous suspension reactors. In particular the Dutch KEMA Suspension Test Reactor (KSTR) was developed, built and successfully operated in the sixties and seventies. It was a 1MWth reactor in which a suspension (6 microns spheres) of mixed UO 2 /ThO 2 in light water was circulated in a closed loop through a sphere-shaped vessel. One of the basic ideas on these suspension reactors was to apply the fission recoil separation effect as a means of purification of the fuel: the non-volatile fission products can be adsorbed in dispersed active charcoal and removed from the liquid. Undoubtedly, this method can present some advantages and better yields for the production of Mo-99 and other short lived radioisotopes, since they have to be extracted from a liquid in which practically no uranium is present. Details are mentioned of the different aspects that have been taken into account and which ones could be added in the corresponding actualization of suspension reactors for radioisotope production. In recent years great advances have been made in nanotechnology that can be used in the tailoring of fuel particles and adsorbent media. Recently, in CNEA Buenos Aires, a new facility has been inaugurated and is being equipped and licensed for laboratory experiments and preparative synthesis of nuclear nanoparticles. RA-6 and RA-3 experimental reactors in Argentina can be used for in-pile testing. (author)

A review of fast reactor program in Japan. April 1997 - March 1998

International Nuclear Information System (INIS)

1998-01-01

This report describes the development and activities on fast reactor in Japan for the period of April 1997 - March 1998. During this period, two important results were drawn by the Special Committee on Fast Breeder Reactors (FBRs) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) Reform Committee, respectively. The Special Committee on FBRs discussed on the future FBR development in Japan including the Prototype FBR 'Monju' operation, and proposed its conclusion as the final report to the Japan Atomic Energy Commission (JAEC) on December 1, 1997. The PNC Reform Committee reviewed PNC's management and safety assurance system, and recommended to reform PNC to a new organization. Each committee result is outlined in this report. The Experimental Fast Reactor 'Joyo' operated 30th - 32nd cycle. In parallel with the operation, the Joyo Upgrading Program (MK-III program) is in progress. Five MK-III driver fuel subassemblies were loaded to the core in the 32nd cycle. Monju comprehensive safety review, which was started in December 1996, was continued through 1997, and was completed in March 1998. The DFBR Plant Optimization (phase 2) design study was launched by the Japan Atomic Power Company (JAPC) with goal of constructing FBR plant that achieves both reliability and economy from FY 1997 for three years. Research and development works are underway under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of PNC, JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). (author)

A review of fast reactor program in Japan. April 2000 - March 2001

International Nuclear Information System (INIS)

Nagata, Takashi; Yamashita, Hidetoshi

2001-01-01

This report describes the development and activities on fast reactors in Japan thru April 2000 to March 2001. During this period, the most important result of the Japanese Fast Reactor Project was that the first phase 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was completed at the end of March 2001, and the second phase study has just started in order to narrow down the candidate concepts selected in the first phase for next stage. In the Experimental Fast Reactor 'Joyo', the 35 th rated power operation was completed by the end of May 2000. The 13 th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) were started on the beginning of June 2000. The modification of the cooling system is underway. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage have already been drawn up based on 'Monju' comprehensive safety review. The Japan Atomic Energy Commission (JAEC) has issued a new 'Long-term Program for Research, Development and Utilization of Nuclear Energy' in November 2000. (author)

Fire safety requirements for electrical cables towards nuclear reactor safety

International Nuclear Information System (INIS)

Raju, M.R.

2002-01-01

Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

Spectrophotometric and pH-Metric Studies of Ce(III, Dy(III, Gd(III,Yb(III and Pr(III Metal Complexes with Rifampicin

Directory of Open Access Journals (Sweden)

A. N. Sonar

2011-01-01

Full Text Available The metal-ligand and proton-ligand stability constant of Ce(III, Dy(III, Gd(III,Yb(III and Pr(III metals with substituted heterocyclic drug (Rifampicin were determined at various ionic strength by pH metric titration. NaClO4 was used to maintain ionic strength of solution. The results obtained were extrapolated to the zero ionic strength using an equation with one individual parameter. The thermodynamic stability constant of the complexes were also calculated. The formation of complexes has been studied by Job’s method. The results obtained were of stability constants by pH metric method is confirmed by Job’s method.

Complexes of lanthanum(III), cerium(III), samarium(III) and dysprosium(III) with substituted piperidines

Energy Technology Data Exchange (ETDEWEB)

Manhas, B S; Trikha, A K; Singh, H; Chander, M

1983-11-01

Complexes of the general formulae M/sub 2/Cl/sub 6/(L)/sub 3/.C/sub 2/H/sub 5/OH and M/sub 2/(NO/sub 3/)/sub 6/(L)/sub 2/.CH/sub 3/OH have been synthesised by the reactions of chlorides and nitrates of La(III), Ce(III), Sm(III) and Dy(III) with 2-methylpiperidine, 3-methylpiperidine and 4-methylpiperidine. These complexes have been characterised on the basis of their elemental analysis, and IR and electronic reflectance spectra. IR spectral data indicate the presence of coordinated ethanol and methanol molecules and bidentate nitrate groups. Coordination numbers of the metal ions vary from 5 to 8. 19 refs.

Problems and prospects of small and medium power reactors

International Nuclear Information System (INIS)

Matin, A.

1977-01-01

Prior to 1973 it was generally believed that small and medium power reactors (SMPRs) had a potentially large market and only their high capital costs prevented their large scale commercial application. In December, 1973, crude oil price rose from US $2.50 per barrel to more than US $11 per barrel. This changed the economic position of SMPRs so much so that even 100-200 MWe nuclear reactors were considered economic compared to oil-fired plants. A Market Survey by the International Atomic Energy Agency in 1974 showed that the potential market for reactors ranging from 150 to 400 MWe during 1980-1990 amounted to 140 units with a total installed capacity of 38,000 MWe. This potential market did not, however, generate the desired interest among the reactor manufacturers. So far only three manufacturers based in Europe have shown interest in SMPRs and at present small reactors are being built commercially only in India. Among developing countries, Bangladesh, Jamaica and Kuwait are seriously looking for reactors in sizes of 100-200 MWe. The paper analyses the historic background of SMPRs and problems related to their commercial application and suggests the following actions: i) The British 100 MWe SGHWR is considered proven and suitable for small grids and hence deserves financial support by British/International Financing Agencies. ii) Any re-engineered or slightly re-designed version of operating small light water reactors will find wider acceptability than available new adaptions of marine reactors. Manufacturers of operating small LWRs may be encouraged through international financial assistance to make such designs commercially available. iii) Small CANDU reactors may be suitable for most developing countries and need technical and economic support from Canada for their export. iv) The Agency must continue their effort more vigorously for making SMPRs commercially available to small developing countries

Fast and sensitive detection of ochratoxin A in red wine by nanoparticle-enhanced SPR.

Science.gov (United States)

Karczmarczyk, Aleksandra; Reiner-Rozman, Ciril; Hageneder, Simone; Dubiak-Szepietowska, Monika; Dostálek, Jakub; Feller, Karl-Heinz

2016-09-21

Herein, we present a fast and sensitive biosensor for detection of Ochratoxin A (OTA) in a red wine that utilizes gold nanoparticle-enhanced surface plasmon resonance (SPR). By combining an indirect competitive inhibition immunoassay and signal enhancement by secondary antibodies conjugated with gold nanoparticles (AuNPs), highly sensitive detection of low molecular weight compounds (such as OTA) was achieved. The reported biosensor allowed for OTA detection at concentrations as low as 0.75 ng mL(-1) and its limit of detection was improved by more than one order of magnitude to 0.068 ng mL(-1) by applying AuNPs as a signal enhancer. The study investigates the interplay of size of AuNPs and affinity of recognition elements affecting the efficiency of the signal amplification strategy based on AuNP. Furthermore, we observed that the presence of polyphenolic compounds in wine samples strongly interferes with the affinity binding on the surface. To overcome this limitation, a simple pre-treatment of the wine sample with the binding agent poly(vinylpyrrolidone) (PVP) was successfully applied. Copyright © 2016 Elsevier B.V. All rights reserved.

Characteristics and economy of the European reactor of pressurized water (EPR)

International Nuclear Information System (INIS)

Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C.

2005-01-01

The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

Dynamic power behavior of a PWR type nuclear reactor

International Nuclear Information System (INIS)

Moreira, F.J.

1984-01-01

A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

Phosphorus removal from UASB reactor effluent by reactive media filtration.

Science.gov (United States)

Rodríguez-Gómez, Raúl; Renman, Gunno

2017-08-01

The phosphorus (P) and BOD7 removal performance of an upflow packed bed reactor (PBR) filled with two reactive filter media was studied over 50 weeks. The lower one-fifth of the reactor was filled with calcium-silicate-hydrate (Sorbulite®) and the upper four-fifth with calcium-silicate (Polonite®). A laboratory-scale upflow anaerobic sludge bed reactor (UASB) delivered wastewater to the PBR. A model was developed to describe the gradient in P concentration change in the reactor, based on reaction kinetics. The reaction terms were assumed to follow the Langmuir isotherm, based on the results obtained in a batch test. First, a comparison was made between experimental and simulated results. The capability of the model to forecast P removal capacity was then tested for three hypothetical cases: (i) reactor filled with Sorbulite and Polonite, (ii) reactor filled with only Sorbulite, and (iii) reactor filled with only Polonite. Finally, a sensitivity analysis was performed for the main parameters in the model. The average removal of P and BOD7 from the UASB effluent was 98% and 90%, respectively. The starting pH of the dual-medium effluent was 12.2 and decreased gradually over time to 11.1. The simulation both overestimated and underestimated mean measured P removal but was within the range of maximum and minimum measured values. The hypothetical cases revealed that most P was removed by Polonite due to calcium phosphate precipitation. The removal capacity of the two filter materials and their layer height in the reactor were the most sensitive parameters in the simulation.

Globally linearized control on diabatic continuous stirred tank reactor: a case study.

Science.gov (United States)

Jana, Amiya Kumar; Samanta, Amar Nath; Ganguly, Saibal

2005-07-01

This paper focuses on the promise of globally linearized control (GLC) structure in the realm of strongly nonlinear reactor system control. The proposed nonlinear control strategy is comprised of: (i) an input-output linearizing state feedback law (transformer), (ii) a state observer, and (iii) an external linear controller. The synthesis of discrete-time GLC controller for single-input single-output diabatic continuous stirred tank reactor (DCSTR) has been studied first, followed by the synthesis of feedforward/feedback controller for the same reactor having dead time in process as well as in disturbance. Subsequently, the multivariable GLC structure has been designed and then applied on multi-input multi-output DCSTR system. The simulation study shows high quality performance of the derived nonlinear controllers. The better-performed GLC in conjunction with reduced-order observer has been compared with the conventional proportional integral controller on the example reactor and superior performance has been achieved by the proposed GLC control scheme.

Development of source term evaluation method for Korean Next Generation Reactor(III)

Energy Technology Data Exchange (ETDEWEB)

Lee, Geon Jae; Park, Jin Baek; Lee, Yeong Il; Song, Min Cheonl; Lee, Ho Jin [Korea Advanced Institue of Science and Technology, Taejon (Korea, Republic of)

1998-06-15

This project had investigated irradiation characteristics of MOX fuel method to predict nuclide concentration at primary and secondary coolant using a core containing 100% of all MOX fuel and development of source term evaluation tool. In this study, several prediction methods of source term are evaluated. Detailed contents of this project are : an evaluation of model for nuclear concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant using purely MOX fuel, suggestion of source term prediction method of NPP with a core using MOX fuel.

Design studies for the Mark-III core of experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

1979-08-01

The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

Concept of passive safe small reactor for distributed energy supply system

International Nuclear Information System (INIS)

Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

2003-01-01

This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

Inner-sphere and outer-sphere complexes of yttrium(III), lanthanum (III), neodymium(III), terbium(III) and thulium(III) with halide ions in N,N-dimethylformamide

International Nuclear Information System (INIS)

Takahashi, Ryouta; Ishiguro, Shin-ichi

1991-01-01

The formation of chloro, bromo and iodo complexes of yttrium(III), and bromo and iodo complexes of lanthanum(III), neodymium(III), terbium(III) and thulium(III) has been studied by precise titration calorimetry in N,N-dimethylformamide (DMF) at 25 o C. The formation of [YCl] 2+ , [YCl 2 ] + , [YCl 3 ] and [YCl 4 ] - , and [MBr] 2+ and [MBr 2 ] + (M = Y, La, Nd, Tb, Tm) was revealed, and their formation constants, enthalpies and entropies were determined. It is found that the formation enthalpies change in the sequence ΔH o (Cl) > ΔH o (l), which is unusual for hard metal (III) ions. This implies that, unlike the chloride ion, the bromide ion forms outer-sphere complexes with the lanthanide(III) and yttrium(III) ions in DMF. Evidence for either an inner- or outer-sphere complex was obtained from 89 Y NMR spectra for Y(ClO 4 ) 3 , YCl 3 and YBr 3 DMF solutions at room temperature. (author)

Behavior of exposed human lymphocytes to a neutron beam of the reactor TRIGA Mark III

International Nuclear Information System (INIS)

Carbajal R, M. I.

2012-01-01

Excessive exposure to ionizing radiation occurs in people who require radiation treatment, also in those for work can come to receive doses above the permitted levels. A third possibility of exposure is the release of radioactive material in which the general population is affected. Most of the time the exhibition is partial and only rarely occurs throughout the body. For various reasons, situations arise where it is impossible to determine by conventional physical methods, the amount of radiation you were exposed to the affected person and in these cases where the option to follow is the Biological Dosimetry, where the analysis of chromosomes dicentrics is used to estimate the dose of ionizing radiation exposure. A calibration curve is generated from in vitro analysis of dicentric chromosome, which are found in human lymphocytes, treated with different types and doses of radiation. The dicentric is formed from two lesions, one on each chromosome and their union results in a structure having two centromeres, acentric fragment with her for the union of several chromosomes leads to more complex structures as tri-centric s, tetra or penta-centric s, which have the same origin. The dose-response curve is estimated by observing the frequency of dicentrics and extrapolated to a dose-effect curve previously established, for which it is necessary that each lab has its own calibration curves, taking into account that for a Let low radiation, dose-effect curve follows a linear-quadratic model Y=C + αD + βD. The production of dicentric chromosomes with a high Let, was studied using a beam of neutrons generated in the reactor TRIGA Mark III with an average energy of 1 MeV, adjusting the linear model Y=αD. The dose-response relationship is established in blood samples from the same donor, the coefficient α of the dose-response is Y = (0.3692 ± 0.011 * D), also shows that saturation is reached in system 4 Gy. (Author)

Liquid plugs bouncing against a solid basis, comparison of SIMMER-III and PLEXUS results

International Nuclear Information System (INIS)

Maschek, W.; Arnecke, G.; Flad, M.

1995-01-01

The SIMMER III code is being tested by application to problems of reactor accidents. The fluid dynamics/thermohydraulics part of the code can be applied also to evaporation/condensation, melt, and freezing phenomena. For a liquid plug bouncing against a solid basis, the momentum transfer is calculated. PLEXUS results turn out to be in significant disagreement with the SIMMER calculations. (orig.)

Investigation for calculation methods used in analyzing the physics characteristics of nuclear power reactor

International Nuclear Information System (INIS)

Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong

2014-01-01

The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)

Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5

International Nuclear Information System (INIS)

Pena V, J.D.

2016-01-01

The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

Program summary for the Civilian Reactor Development Program

International Nuclear Information System (INIS)

1982-07-01

This Civilian Reactor Development Program document has the prime purpose of summarizing the technical programs supported by the FY 1983 budget request. This section provides a statement of the overall program objectives and a general program overview. Section II presents the technical programs in a format intended to show logical technical interrelationships, and does not necessarily follow the structure of the formal budget presentation. Section III presents the technical organization and management structure of the program

Safety research for CANDU reactors

International Nuclear Information System (INIS)

Hancox, W.T.

1982-10-01

Continuing research to develop and verify computer models of CANDU-PHW reactor process and safety systems is described. It is focussed on loss-of-coolant accidents (LOCAs) because they are the precursors of more serious accidents. Research topics include: (i) fluid-dynamic and heat-transfer processes in the heat transport system during the blowdown and refilling phases of LOCAs; (ii) thermal and mechanical behaviour of fuel elements; (iii) thermal and mechanical behaviour of the fuel and the fuel-channel assembly in situations where the heavy-water moderator is the sink for decay heat produced in the fuel; (iv) chemical behaviour of fission gases that might be released into the reactor coolant and transported to the containment system; and (v) combustion of hydrogen-air-steam mixtures that would be produced if fuel temperatures were sufficiently high to initiate the zirconium-water reaction. The current status of the research on each of these topics is highlighted with particular emphasis on the conclusions reached to date and their impact on the continuing program

Analysis and application of a simulator of a nuclear reactor AP-600

International Nuclear Information System (INIS)

Medina S, V. S.; Salazar S, E.

2011-11-01

In front of the resurgence of interest in the nuclear power production, several national organizations have considered convenient to have highly specialized human resources in the technologies of nuclear reactors of III + and IV generation. For this task, the intensive and extensive applications of the computation should been considered, as the virtual instrumentation. The present work analyzes the possible applications of a nuclear simulator provided by the IAEA with base in the design of the reactor AP-600, using a focusing of modular model developed in FORTRAN. One part of the work that was made with the simulator includes the evaluation of 21 transitory events of operation, including the recreation of the accident happened in the nuclear power plant of Three Mile Island in 1979, comparing the actions flow and the answer of the systems under the intrinsic security of a III + generation reactor. The impact that had the mentioned accident was analyzed in the growing of the nuclear energy sector and in the public image with regard to the nuclear power plants. An application for this simulator was proposed, its use as tool for the instruction in the nuclear engineering courses using it to observe the operation of the different security systems and its interrelation inside the power plant as well as a theoretical/practical approach for the student. (Author)

Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

International Nuclear Information System (INIS)

Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

2004-01-01

Full text: The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by (1) implementing safety upgrades, or (2) assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

International Nuclear Information System (INIS)

Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

2004-01-01

The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by implementing safety upgrades, or assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

The ARIES-II and ARIES-IV second-stability tokamak reactors

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.; Hasan, M.Z.; Mau, T.-K.; Sharafat, S.; Baxi, C.B.; Leuer, J.A.; McQuillan, B.W.; Puhn, F.A.; Schultz, K.R.; Wong, C.P.C.; Brooks, J.; Ehst, D.A.; Hassanein, A.; Hua, T.; Hull, A.; Mattis, R.; Picologlou, B.; Sze, D.-K.; Dolan, T.J.; Herring, J.S.; Bathke, C.G.; Krakowski, R.A.; Werley, K.A.; Bromberg, L.; Schultz, J.; Davis, F.; Holmes, J.A.; Lousteau, D.C.; Strickler, D.J.; Jardin, S.C.; Kessel, C.; Snead, L.; Steiner, D.; Valenti, M.; El-Guebaly, L.A.; Emmert, G.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.; Sviatoslavsky, I.N.; Cheng, E.T.

1992-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currently planned for the ARIES program. The ARIES-I 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. The ARIES-III study focuses on the potential of tokamaks to operate with D- 3 He fuel system as an alternative to deuterium and tritium. The ARIES-II and ARIES-IV designs have the same fusion plasma but different fusion-power-core designs. The ARIES-II reactor uses liquid lithium as the coolant and tritium breeder and vanadium alloy as the structural material in order to study the potential of low-activation metallic blankets. The ARIES-IV reactor uses helium as the coolant, a solid tritium-breeding material, and silicon carbide composite as the structural material in order to achieve the safety and environmental characteristic of fusion. In this paper the authors describe the trade-off leading to the optimum regime of operation for the ARIES-II and ARIES-IV second-stability reactors and review the engineering design of the fusion power cores

The Fort St. Vrain high temperature gas-cooled reactor. III

International Nuclear Information System (INIS)