Architectural and town-planning reconstruction problems of the city of Voronezh

Science.gov (United States)

Mikhaylova, TTatyana; Parshin, Dmitriy; Shoshinov, Vitaly; Trebukhin, Anatoliy

2018-03-01

The analysis of the state of the historically developed urban district of the city of Voronezh is made. The ways of solving the identified architectural and urban problems of reconstruction of historically developed buildings are proposed. The concept of reconstruction of a territory with historical buildings along Vaytsekhovsky Street is presented.

Trends in the physical development indicators of 14-year-old adolescents in the Voronezh Region

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V. P. Sitnikova

2014-01-01

Full Text Available The paper presents the results of statistical (sigma and percentile analysis of height and weight in 434 boys and girls aged 14 years from the Voronezh Region in 2011 —2012. The girls' height was comparable with the regional indicators of physical development in the children of the Voronezh Region (1997—1999. The boys' height was characterized by a wide scatter of the obtained values; but there was an increase in the mean values as compared to the 1997—1999 data. Examinations of the boys and girls revealed a wide scatter in their weight with its increasing tendency. The performed examinations indicate the need for revision of the regional height and weight values for teenagers every ten years.

Monitoring of Ecological and Geochemical State of the Soil Cover in the City of Voronezh

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Sereda Lyudmila Olegovna

2015-09-01

Full Text Available Soil cover in the city of Voronezh accumulates a lot of pollutants and indicates the centers of technological pollution. The high rates of housing construction, functioning and development of urban infrastructure cause infringement to the soil cover. The paper contains main results of an ecological and geochemical research of the soil cover in Voronezh, its characteristics, properties of the horizons of the different types of soils. During spring and summer of 2014 75 samples of soil were collected in special points of monitoring (according to GOST 17.4.3.01-83 and GOST 17.4.4.02-84. During the research the following methods were applied – volt-ampermetric method was used for detecting the concentration of heavy metals, the method of cholophorm-hexan extraction – for petrochemicals, the method of I.V. Tyurin – for humus concentration, potentiometric method and biotesting methods (analysis of seedlings of the following indicating plants – Lepidium sativum, Avena sativa, as well as defining the phytotoxic effect – for actual acidity detection. The obtained results are used for creating an overview soil map of Voronezh. Urbanozems are dominating in the soil cover of Voronezh. There era large areas of them in the majority of the city districts. A smaller part of a total urban area is presented by soils, which are slightly touched by human economic activity. Urban soils of industrial and transport city zones have disadvantageous properties – low rate of humus and alkali reaction of soil environment, high rate of pollution by petrochemicals and heavy metals. The least rate of pollution of a soil cover by heavy metals is detected in residential areas, situated far from industrial objects and highways. We have detected dependence between accumulation of polluting substances in soil cover and functional and planning peculiarities of the city. For example, accumulation of zinc takes place in soils with alkali reaction of soil and low

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

International Nuclear Information System (INIS)

Yang, Changkeun; Kim, Yohan; Ha, Sangjun

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Aspartate aminotransferase (AST) blood test

Science.gov (United States)

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

The Features of Naturalization of Invasive Fraction of Flora in the Voronezh Region and in Some Regions of the European Part of Russia

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Vladimirov D.R.

2015-10-01

Full Text Available The article is about naturalization features of invasive fraction of flora in Voronezh and some other regions of the European part of Russia. The summary table represents all invasive and potentially invasive plants of the European part of Russia with their level of naturalization (or invasive status. Invasive fraction of flora in the Voronezh region numbers 120 plants. All of them are on different stages of naturalization process in an anthropogenic areal. Invasive plants represent by agriophyts – 41 (34,1 % species, epecophyts – 75 (62,5 % species and colonophyts-epecophyts – 4 (3,4 % species. Totally there are 201 species of invasive and potentially invasive plants spread within European part of Russia (Northern-West Russia, Ivanovo, Kaluga, Tver, and Voronezh regions. They formed the “black list” of European Russia. 10 species are common to all invasive fractions. These are Acer negundo, Amelanchier x spicata, Aster x salignus, Echinocystis lobata, Elodea canadensis, Heracleum sosnowskyi, Impatiens glandulifera, Impatiens parviflora, Juncus tenuis and Lupinus polyphyllus. The analysis of the general list of invasive fractions of European Russia shows that 120 species of the list are invasive or potentially invasive in the Voronezh region (100 and 20 species in accordance, adventives naturalized species – 31, native species – 19, archaeophyts – 2, apophyts – 4. 26 species from the list were not found in the Voronezh region. Apparently, the region is a transit area for many invasive plants, which migrate from South to North, from East to West etc. Not only its natural and climatic potential, but also high level of transformation of local landscapes enabled immigrant-plants to naturalize within the bounds of the region. Furthermore, for many years the Voronezh region was the center of introduction of alien plants. Many of those became a part of invasion fraction of regional flora. In recent decades green building took place of

The role of small and medium-sized high-tech enterprise in the formation of innovative potential of the Voronezh region

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E. A. Shubina

2016-01-01

Full Text Available In modern conditions the region's sustainable development and economic growth is largely determined by its ability to implement innovations. One of the determining factors of long-term economic development of the territories, to date, is innovation. A mechanism to give impetus to the innovative development of the regions is to create effective small and medium-sized innovative enterprises. Voronezh region, like many Russian regions, has considerable potential for innovation, but usually it is not used efficiently or not used at all. The article deals with the concept and the need to develop innovative entrepreneurship and the role of small and medium innovative enterprises in the formation of scientific and technical potential of the Voronezh region. The basic conditions and factors promoting and impeding the development of small and medium innovative businesses in the region, determined the willingness of subjects of innovation activity in the development, implementation and promotion of innovation in the region. The problems of development of small and medium-sized businesses at the regional level. The evaluation of actions to improve SMEs management programs in the region. Particular attention is given to the classification of small and medium-sized innovative business structures. In order to stimulate the development of small and medium-sized innovative business in the Voronezh region in the article suggests the implementation of a package of measures. The active use of innovation by small and medium-sized enterprises of the Voronezh area enhances their efficiency and competitiveness, creates new jobs, which ultimately has a positive effect on the region's economic development, the growth of the tax base, improving the quality of life of the population.

JUSTIFICATION OF THE PRIORITIES OF THE CLUSTERING OF AGRO-INDUSTRIES OF THE VORONEZH REGION

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Y. A. Salikov

2014-01-01

Full Text Available Currently, in many regions of the Russian Federation initiated a large-scale work on the development and implementation of cluster policy in accordance with Federal and regional socio-economic development until 2020. The analysis of the status of implementation adopted in 2012, the concept of cluster policy of the Voronezh region showed that the complex is made on the date of the event is mainly responsible for the informational and infrastructural nature. However, from the total number of promising clusters by 2014, formed in fact, only two-thirds, while among the uncreated shall apply the cluster processing of agricultural products having a high rating prospects. Given that the formation of the agro-industrial cluster corresponds to the requirements and conditions in this study developed a new methodological approach, which carried out the rationale for the priority of the formation of the meat cluster in the agro-industrial complex of the Voronezh region. The basis of this methodological approach is the algorithm for the identification of areas of clustering, developed by the authors using statistics Forsythe, represents an efficient tool for the formation of priorities to achieve a qualitatively new results in the field of economy, science and technology. The proposed algorithm includes the serial combination of the following methodological stages: the formation of the object of research, identifying sources of reliable information on the basis of expert assessments, identify areas clustering of industries (including analysis legal framework the study of statistical data on the level of localization of industries and analysis of the practice of implementation of the cluster policy regions-analogues, identification of areas for additional clustering of industries and their mapping, and de-termination of the priority directions of the additional clustering of industries by ranking. The results of the study, carried out in accordance with this

AST Test: MedlinePlus Lab Test Information

Science.gov (United States)

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

The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

International Nuclear Information System (INIS)

Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

1993-01-01

Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

PhAST: pharmacophore alignment search tool.

Science.gov (United States)

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

2009-04-15

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

Asteroids Dynamic Site-AstDyS

Science.gov (United States)

Knezevic, Zoran; Milani, Andrea

2012-08-01

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

Estimation of the condition of snow cover in Voronezh according to the chemical analysis of water from melted snow

OpenAIRE

Prozhorina Tatyana Ivanovna; Bespalova Elena Vladimirovna; Yakunina Nadezhda

2014-01-01

Snow cover possesses high sorption ability and represents informative object to identify technogenic pollution of an urban environment. In this article the investigation data of a chemical composition of snow fallen in Voronezh during the winter period of 2014 are given. Relationships between existence of pollutants in snow and the level of technogenic effect are analyzed.

Cohesion assessment of student groups from the faculty of dentistry at Burdenko Voronezh State Medical University

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Pashkov A. N.

2017-10-01

Full Text Available the article presents data on the study of the cohesion of 115 students from the faculty of dentistry at Burdenko Voronezh State Medical University. The study revealed that students have high and medium favorable psychological climate. 99 of them have the average level of group cohesion, and 16 revealed a low level of this indicator. To improve the educational process and interpersonal relations, these results must be taken into account when forming groups.

[Evaluation of modern epizootic activity of natural tularemia foci in Voronezh region using immune-serological and molecular-genetic study of main carriers of the disease].

Science.gov (United States)

Meshcheriakova, I S; Trankvilevskiĭ, D V; Kvasov, D A; Mikhaĭlova, T V; Kormilitsina, M I; Demidova, T N; Stepkin, Iu I; Zhukov, V I

2015-01-01

Improvement of monitoring and prognosis of epidemic manifestations of natural foci of tularemia on the territory of Voronezh region using immune-serological and molecular-genetic study of main carriers of the disease. 539 small mammals captured during summer period of 2011 in 4 districts of North-Eastern part of Voronezh region were studied. Animal organs were studied by serologic (search for Francisella tularensis antigens) and molecular-biologic (detection of F. tularensis DNA) methods. Tularemia antigen was detected using passive hemagglutination reaction (PHAR) with erythrocytic tularemia immunoglobulin diagnosticum. Real-time polymerase chain reaction (RT-PCR) was applied for detection of tularemia causative agent DNA. Complex study revealed epizootic activity of natural foci of tularemia in the examined territory. F. tularensis antigen and/or DNA were detected in 82 objects (15.2%). Use of RT-PCR allowed to additionally detect samples with relatively low content of F. tularensis DNA substrate, when antigen was not detected in samples. High sensitivity and specificity of the RT-PCR was ensured by inclusion of specific probes (tu14-PR2 and ISFTu2P). The results obtained give evidence on functioning and epizootic activity of natural foci of tularemia in Voronezh region that requires constant monitoring of the territory and prophylaxis measures, first of all vaccination of risk groups by live tularemia vaccine.

Additional data to the stratigraphy and the chronology of the Kostenki 1 (Poliakov) sequence, Voronezh, Russia : Le Sungirien, Saint-Petersbourg 2016

NARCIS (Netherlands)

Haesaerts, P; Damblon, F; van der Plicht, Johannes; Otte, Marcel; Nigst, Phillip R.

2017-01-01

Kostenki 1 is one of the many sites of the Kostenki- Borshchevo site cluster south of Voronezh, which has a long sequence covering the Early and Mid Upper Palaeo- lithic, including the Streletskian Cultural Layer V. Here we present stratigraphic data from our 1994 eldwork (sections of the 1981-1982

The experience of functioning of the dairy cluster of the Voronezh region in the aspect of observing the criteria of economic efficiency

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A. V. Kotarev

2018-01-01

Full Text Available In the article the analysis of tendencies of development of sector of dairy animal industries on the basis of formation of a dairy cluster is given. It is shown that the structure of dairy cattle breeding in the Voronezh region is built in the form of a milk cluster, in the functioning of which most of the existing enterprises of the dairy industry take part. The largest participants of the dairy cluster are: EkoNivaAgro LLC, GK Molvest, Agroholding Don-Agro. In the cluster, all the operations of the whole milk production process are carried out, starting from the production of feed and breeding of breeding animals and ending with the delivery of finished products to the end users. Also, the cluster included auxiliary organizations that carry out scientific, educational, research and veterinary support of the entire production chain. Within the framework of the cluster, effective interaction of all its participants is established, which is necessary for obtaining high quality milk and dairy products. The advantage of a dairy cluster in the Voronezh region is the opportunity to provide targeted benefits to organizations that are important for the economy of the region, and the state gets the opportunity to regulate innovation, investment flows and evaluate the effectiveness of financial investments. The implementation of investment projects in animal husbandry in the medium term (for the period until 2020 will increase milk production by 30%. The economic efficiency of dairy cattle is a cumulative category that reflects the impact of technological, economic, social and environmental factors on the productivity of production. Cluster development of the dairy cattle breeding industry in the Voronezh Region allows obtaining a positive synergistic effect, reducing transaction costs, and increasing the genetic potential of dairy cattle.

Aerotechnogenic Monitoring of Urban Environment on Snow Cover Pollution (on the Example of Voronezh City

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Prozhorina Tatyana Ivanovna

2014-09-01

Full Text Available Snow cover is characterized by high sorption ability and represents an informative object in the process of identifying the technogenic pollution of urban environment. The article contains the results of the research on the chemical composition of the snow which fell in Voronezh in the winter period of 2013–2014. The coefficients of chemical elements concentration were calculated to provide objective characteristics of snow cover pollution. The authors analyze the connection between the presence of pollutants in snow and the level of technogenic impact. The obtained ranges of anomaly coefficients among anions reflect the composition of technogenic emissions. The mineralization of snow water reliably characterizes the intensity of anthropogenic impact on the urban environment, and the value of mineralization snow samples ranges from 62,6 (background to 183,9 mg/l. Maximum values of mineralization (more than 150 mg/l are typical for samples taken in transport area. High values of salinity (more than 120 mg/l are observed in snow samples taken in the industrial area, which confirms the high “technogenic pressure” on the urban environment in zones of industrial and transport potential of the city. The investigated functional areas can be arranged in the following series by descending level of contamination: transport area > industrial zone > residential and recreational areas > background territory. The study of the chemical composition of snow cover in the various functional areas of Voronezh allows to conclude that the pH level, mineralization and the content of suspended solids in snow waters characterize the intensity of anthropogenic pressure on the urban environment, and the composition of melt waters indicates the nature of its pollution.

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

International Nuclear Information System (INIS)

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

2010-01-01

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

A particle bed reactor based NTP in the 112,500 N thrust class

International Nuclear Information System (INIS)

Ludewig, H.; Powell, J.R.; Lazareth, O.W. Jr.; Todosow, M.

1993-01-01

This paper discusses the application of a Particle Bed Reactor (PBR) to a 112,500 N thrust Nuclear Thermal Propulsion (NTP) Engine. The method of analysis is described, followed by a presentation of the results. It is concluded that the PBR would result in a very competitive NTP engine. In addition, due to the high power densities possible with a PBR, high thrust/weight ratios are possible. This conclusion can be used to satisfy a variety of mission goals

A particle bed reactor based NTP in the 112,500 N thrust class

Science.gov (United States)

Ludewig, Hans; Powell, James R.; Lazareth, Otto W.; Todosow, Michael

1993-01-01

This paper discusses the application of a Particle Bed Reactor (PBR) to a 112,500 N thrust Nuclear Thermal Propulsion (NTP) Engine. The method of analysis is described, followed by a presentation of the results. It is concluded that the PBR would result in a very competitive NTP engine. In addition, due to the high power densities possible with a PBR, high thrust/weight ratios are possible. This conclusion can be used to satisfy a variety of mission goals.

AstRoMap European Astrobiology Roadmap.

Science.gov (United States)

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

2016-03-01

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

ELEVATED ALT AND AST IN AN ASYMPTOMATIC PERSON

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KEW ST

2009-01-01

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

Design and construction of a 7,500 liter immobilized cell reactor-separator for ethanol production from whey

Energy Technology Data Exchange (ETDEWEB)

Dale, M.C.

1992-12-31

A 7,500 liter reactor/separator has been constructed for the production of ethanol from concentrated whey permeate. This unit is sited in Hopkinton IA, across the street from a whey generating cheese plant A two phase construction project consisting of (1) building and testing a reactor/separator with a solvent absorber in a single unified housing, and (2) building and testing an extractive distillation/product stripper for the recovery of anhydrous ethanol is under way. The design capacity of this unit is 250,000 gal/yr of anhydrous product. Design and construction details of the reactor/absorber separator are given, and design parameters for the extractive distillation system are described.

Research works at the Physics Institute nuclear reactor for the nuclear power engineering

International Nuclear Information System (INIS)

Gavars, V.V.; Kalnin'sh, D.O.; Lapenas, A.A.; Tomsons, E.Ya.; Ulmanis, U.A.

1985-01-01

Methods for neutron spectra determination in the nuclear reactor core and vessel have been developed. On their base the neutron spectra at the Novo-Voronezh and kola NPPs have been measured. Such measurements are necessary for the determination of the nuclear fuel reprocessing coefficients, for the evaluation of the construction radiation-damage stability and the NPP economical efficiency on the whole. A new type of the reactor regulator - a liquid metal one - has been created. Such regulators are promising in respect of their use at the NPPs. The base for studying new radiation-damage-stable insulators has been founded. The materials obtained are now applied to designing the reactors of the second (fast) and the third (thermonuclear H) generations. There have developed and by a long-time exploitation checked a hot loop, used for materials irradiation. the nuclear reactor in Salaspils provides training of students being the new brain-power for the nuclear power engineering

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

Science.gov (United States)

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

2016-10-01

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

Passive safety systems for integral reactors

International Nuclear Information System (INIS)

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

1996-01-01

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

Passive safety systems for integral reactors

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

Concept and designs of new-generation fast reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.

1993-01-01

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

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

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K. Nsiah

2011-01-01

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

A new method for synthesis of As-Te chalcogenide films

Science.gov (United States)

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

2017-11-01

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

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

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Yuko Inami

2014-01-01

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

Ecological Resource and its realization in the Botanical Garden of the Voronezh State University

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Voronin Andrey Alekseevich

2016-12-01

Full Text Available The article considers components of environmental resource of the Botanical Garden of Voronezh State University in connection with its introductional activity and scientific and educational process. The components include collections, expositions, area of natural vegetation typical of forest-steppe zone. The article provides examples of ecological resource use that is study of vascular plants flora, mosses, lichens, fungi, fauna and of ecological and biological characteristics of collection plants, with priority for rare and endangered species. There is also description of field training and work practices, excursions, talks. Special attention is given to Botanical Garden potential in tourist and excursion arrangement as well as development of ecological vision. 13 items of environmental and educational path were identified and described which include all natural ecosystems and old dendrology collections, fallow lands. The excursions themes for all other collections of the Botanical Garden were named. Real perspectives for science research were defined.

Accident transient processes at NPPs with the WWER type reactors

International Nuclear Information System (INIS)

Bukrinskij, A.M.

1982-01-01

Thermal-physical and nuclear-physical transient processes at NPPs with the WWER type reactors during accidents with the main technological equipment failures and the accidents with loss of coolant in the primary and secondary coolant circuits are considered. Mathematical methods used for these processes modelling is described. Examples of concrete calculations for accidents with different failures are given. Comparative analysis of the results of dynamic tests at the Novo-Voronezh-3 reactor is presented. It is concluded that the modern NPP design is impossible without application of mathematical modelling methods. The mathematical modelling of transients is also necessary for proper and safe NPP operation. Mathematical modelling of accidents at NPPs is a comparatively new method of investigation. Its success and development are completely based on the progress in modern computer development. With their improvement the mathematical models will become more complicate and adequacy of real physical process representation by their means will increase

Kirjavahetus Kaarel Robert Pustaga / Karl Ast Rumor

Index Scriptorium Estoniae

Rumor, Karl, pseud., 1886-1971

2009-01-01

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

Thermal-hydraulic design of the 200 MW NHR

Energy Technology Data Exchange (ETDEWEB)

Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

1997-09-01

The main problems regarding the AST-500 NHR thermal-hydraulics are considered. Basic thermal data of the reactor plant are given and peculiarities of coolant parameters at natural convection in the primary circuit are discussed. The in-reactor instrumentation system is briefly describes, as well as the results of natural-convective flow characteristics investigations using reactor test models. (author). 4 refs, 5 figs.

Review of NHR activities in the Russian Federation

International Nuclear Information System (INIS)

Malamud, V.A.; Kurachenkov, A.V.; Kusmartsev, E.V.

1997-01-01

NHR development activities in the ex-USSR were initiated in the 1970s mainly due to a growing deficiency of organic fuels needed for heating large cities in the European part of the country. Construction of two pilot nuclear district heating plants with AST-500 NHRs was started in the early 1980s, and by 1989 the first unit in Gorky NDHP was nearly 90% completed. Current activity in this field is concentrated on upgrading the AST-500 design and on the development on this basis of a whole series of heating-only and co-generation reactor plants of unit power ranging from 30 to 600 MW. A brief description of the AST-500 reference NHR design features is given, as well as of the R and D activities that have been carried out for the design decisions and safety validation. (author). 12 refs, 1 tab

Review of NHR activities in the Russian Federation

Energy Technology Data Exchange (ETDEWEB)

Malamud, V A; Kurachenkov, A V; Kusmartsev, E V [OKBM, Nizhny Novgorod (Russian Federation)

1997-09-01

NHR development activities in the ex-USSR were initiated in the 1970s mainly due to a growing deficiency of organic fuels needed for heating large cities in the European part of the country. Construction of two pilot nuclear district heating plants with AST-500 NHRs was started in the early 1980s, and by 1989 the first unit in Gorky NDHP was nearly 90% completed. Current activity in this field is concentrated on upgrading the AST-500 design and on the development on this basis of a whole series of heating-only and co-generation reactor plants of unit power ranging from 30 to 600 MW. A brief description of the AST-500 reference NHR design features is given, as well as of the R and D activities that have been carried out for the design decisions and safety validation. (author). 12 refs, 1 tab.

AGS Spallation Target Experiment (ASTE) Collaboration

International Nuclear Information System (INIS)

Oyama, Yukio

1999-01-01

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

Irradiations under magnetic field. Measurement of resistivity sample irradiations between 100 and 500 deg C in a swimming-pool reactor

International Nuclear Information System (INIS)

Pauleve, J.; Marchand, A.; Blaise, A.

1964-01-01

An oven is described which enables the irradiation of small samples in the maximum neutron flux of a swimming-pool reactor of 15 MW (Siloe), at temperatures of between 100 and 500 deg.C defined to ± 0,5 deg.C, The oven is very simple from the technological point of view, and has a diameter of only 27 mm, This permits resistivity measurements to be carried out under irradiation in the reactor, or as another example, it enables irradiations in a magnetic field of 5000 oersteds, created by an immersed solenoid. (authors) [fr

Investigations on accidents with massive water ingress exemplified by the pebble bed reactor PNP-500

International Nuclear Information System (INIS)

Moormann, R.

1986-01-01

A computer code is used for analyses of massive water ingress accidents in the High-Temperature Gas Cooled Reactor concept PNP-500 with pebble bed core. The analyses are mainly focussed on graphite corrosion processes. For the investigated accidents a correct reactor shut down in assumed. The mass of water ingressing into the primary circuit is varied between 1000 and 7500 kg (i.e., up to hypothetical values). The dependence of accident consequences on parameters such as intensity and starting time of the afterheat removal system or kinetic values of the chemical processes is examined. The results show that even under pessimistic assumptions the extent of the graphite corrosion is relatively low; significant damaging of fuel elements or graphite components does not occur. A primary circuit depressurization, combined with local burning of water gas, would probably not affect the fission product retention potential of the (gastight) containment. Summing up, the risk caused by these accidents remains small. (orig.) [de

Experimental investigations on the transient behaviour of nuclear heat plants with natural convection

International Nuclear Information System (INIS)

Adam, E.; Sydow, J.; Wolff, J.

1988-01-01

Apart from the theoretical approach, practical experiments concerning the transient behaviour of the primary loop of reactors with natural coolant convection are necessary in order to evaluate the safety systems of reactors providing heat for industrial and communal consumers. The article presents experiments concerning the transient behaviour of the experimental plant DANTON, which models the reactor AST-500, and gives a preview of further research. (orig.) [de

Status of the HTR 500 design program

International Nuclear Information System (INIS)

Baust, E.; Arndt, E.

1987-01-01

Since 1982 BBC/HRB have offered the HTR 500 as the follow-on project of the THTR 300, the first large pebble bed reactor. The technical concept of the HTR-500 largely corresponds to the THTR 300 which has been in operation for almost 2 years now. In developing the design concept of the HTR 500 the ideas and demands of the reactor users in the FRG interested in the HTR have been taken into consideration to a large extent. In 1982 these potential users formed a working group 'Arbeitsgemeinschaft Hochtemperaturreaktor' (AHR), representing 16 power indusry companies and in early 1983, awarded a contract to HRB to perform a conceptual design study on the HTR 500. Within this conceptual design study BBC/HRB developed the safety concept of the HTR 500, prepared a detailed description of the overall power plant, and performed a cost calculation. These activities were completed in 1984. Based on the positive results of this conceptual design study, BBC/HRB are expecting to be granted a design contract by the users company Hochtemperaturreaktor GmbH (HRG) to establish the final complete design plans and documents for the HTR 500. (author)

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

OpenAIRE

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

2016-01-01

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

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

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Shingo Hatakeyama

2012-01-01

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

Model prototype of information support system for operator approaches and realization

International Nuclear Information System (INIS)

Samoilov, O.B.; Galushkin, V.A.; Drumov, V.V.; Kurachenkov, A.V.; Shashkin, S.L.; Mordvincev, V.M.

1998-01-01

In connection with the appearance in a structure of the national regulatory documentation on safety of the requirement about availability of information support systems, the works on development of such system are necessary for making a decision of a question on start-up each NPP. It was developed a model - prototype of the system for Voronezh AST (VAST). Main principles of this model are described in the present report. Besides, similar works on other types of NPPs are carried out. (author)

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

Science.gov (United States)

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

2018-04-01

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

Phosphatized algal-bacterial assemblages in Late Cretaceous phosphorites of the Voronezh Anteclise

Science.gov (United States)

Maleonkina, Svetlana Y.

2003-01-01

Late Cretaceous phosphogenesis of the Voronezh Anteclise has occurred during Cenomanian and Early Campanian. SEM studies show the presence of phosphatized algal-bacterial assemblages both in Cenomanian and Campanian phosphorites. In some Cenomanian nodular phosphorite samples revealed empty tubes 1 - 5 microns in diameter, which are most likely trichomes of cyanobacterial filaments. Other samples contained accumulations of spheres 0,5-3 microns, similar to coccoidal bacteria. Complicated tubular forms with variable diameter 2 - 5 microns occur on surface of some quartz grains in nodules. They are probably pseudomorphs after algae. We found similar formations in the Campanian phosphate grains. Frequently, grain represents a cyanobacterial mat, which is sometimes concentrically coated by phosphatic films. The films of some grains retain the primary structure, their concentric layers are formed by pseudomorphs after different bacterial types and obviously they represent oncolite. In other cases, the primary structure is unobservable because of recrystallization process erases them. Occasionally, the central part retains the coccoidal structure and the recrystallization affects only films. Besides the core of such oncolite can be represented not only by phosphatic grain, but also by grains of other minerals, such as quartz, glauconite and heavy minerals, which serve as a substrate for cyanobacterial colonies. Bacteria also could settle on cavity surfaces and interiors frames of sponge fragments, teeth and bones.

INTEGRATED ASSESSMENT AND GISMAPPING OF THE ENVIRONMENTAL STATE OF THE CITY OF VORONEZH (RUSSIA

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Semen A. Kurolap

2015-01-01

Full Text Available The authors have created a geoinformation-analytical system (GIS for integratedassessment and mapping of the ecological conditions of the territory according to the criteria of anthropogenic impact and quality of the urban environment, as well as the response of woody plants and the health of the child population (on the example of Voronezh – the largest industrial city of the Central Chernozem region.It has been identified that anthropogenic pollution is formed by the industrial-transportsector and varies with regard to the features of the functional planning infrastructure; near the industrial facilities of the petrochemical profile in the left-Bank sector of the city, conditions for the existence of woody plants significantly worsen, which is manifested in the inhibition of their development; child morbidity rate is significantly higher in industrially polluted neighborhoods with high load of pollutant emissions from industry and transport. The diseases primarily associated with pollution are congenital anomalies, neoplasms, endocrine pathology and diseases of the urogenital area.The industrial zone is the main contributor to the total pollution of air, but the transport zoneis the main contributor to the total pollution of soil and snow cover.

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

Science.gov (United States)

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

2016-01-01

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

FEM analysis of foundation raft for 500 MWe pressurized heavy water reactor building

International Nuclear Information System (INIS)

Kulkarni, N.N.; Goray, J.S.; Joshi, M.H.; Paramasivam, V.

1989-01-01

Foundation raft supports the containment structure and internals for 500 MWe PHW reactor building. It also serves as bottom envelop of the containment structure. In view of this, the design of foundation raft assumes great importance. The foundation raft is subjected to various load, most significant of them are dead load of structure, equipment loads transferred through a system of floors, walls and structural steel columns, pressure load during accident conditions, seismic loads, earth pressure, uplift due to buoyancy loads, foundation reaction etc. In order to achieve optimum design, the detailed structural analysis is required to be performed methodically and in most realistic manner. Finite element methods which have come in vogue with the developments in digital computers can be successfully applied in this area. The paper describes the above methods in detail for the analysis of foundation raft for the various load combinations required to be considered for safe and optimum design

The safety characteristics of the HTR 500 reactor plant

International Nuclear Information System (INIS)

Wachholz, W.

1987-01-01

The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

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

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Shah Md Fazlul Karim

2015-03-01

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

Stability characteristics of the 500 mw Indian PFBR

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Anuraj Vijayan L.

2015-01-01

Full Text Available After the successful operation of the fast breeder test reactor for over two decades, India is now nearing the completion of a 500 MW (electrical prototype fast breeder reactor. This commercial scale power reactor is a sodium-cooled, pool-type, mixed-oxide fuelled fast reactor. The stability characteristics of the reactor are an important safety aspect to be studied. In the present work, linear stability of the prototype fast breeder reactor analysis is carried out using the transfer function method, while the stability of the system is checked via the Nyquist criteria. For the completeness of the study, transient analysis with various kinds of reactivity perturbations was carried out. The response of the system in both cases indicated that the system is stable.

The analysis of consumer preferences residents of Voronezh in respect of products of functional purpose from fruit and berry raw materials

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I. P. Shchetilina

2016-01-01

Full Text Available The last decade, the negative trend related to the health status of the population. In the structure of power a significant part of the population are violations. According to scientific research Institute of nutrition in our country commonplace latent forms of vitamin deficiency combined with lack of calcium, iron, iodine. With the aim of increasing the production of food products of mass consumption enriched with minerals and vitamins, the use of vegetable raw materials of the Voronezh region is a priority. The human body for normal functioning requires regular consumption of micronutrients. An important role in biological processes in which food is converted into energy, play a micro- and macroelements and vitamins. They provide the protective functions of the body are involved in tissue renewal. However, micronutrients are not synthesized in the body, and water-soluble vitamins, unlike fat-soluble are easily excreted from the body, so they should regularly come directly from food. In the works of foreign and domestic scientists the development of product functionality has received considerable attention. Special contribution study of academicians V. M. Bosnakovski, V. A. Tutelyan, professor B. P. Sukhanov and others In the development of new types of foods must take into account the views of consumers about new products. The paper presents marketing research of the market of Voronezh the survey. The aim of the study was to identify preferences of consumers of functional products. Analyzed socio-demographic profile of respondents the distribution of respondents by education, age of the interviewee, the frequency of consumption by respondents functional products, the preferences of respondents by frequency of consumption depending on the sex of the respondents. The analysis of preferences for specific groups of products, places to purchase products functional purpose, the reasons that motivate respondents to purchase products of functional

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

Science.gov (United States)

Baggio, Gustavo Inheta; Hage, Simone Rocha de Vasconcellos

2017-12-07

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

Experience in lifetime extension of the first generation WWER-440 power units

International Nuclear Information System (INIS)

Medvedev, P.

2002-01-01

In connection with the expiration of the lifetime for the first generation WWER-440 reactors in Russian Federation (Novo Voronezh and Kola NPP), the legal procedures and Life Time Extension (LTE) Program are discussed. The LTE Program includes: development of regulation basis; economic efficiency studies; power unit modernization; power unit comprehensive examination and justification od equipment resource; in-dept safety assessment; operational license acquisition. As a result from the LTE Program the safety level of the unit 3 of the Novo Voronezh NPP is significantly increases, the operational period has been justified and a 5-year license has been issued

Atomic science and engineering in the economy of the Soviet Union

International Nuclear Information System (INIS)

Kruglov, A.K.

1976-01-01

The main achievements of Soviet atomic science and engineering are presented. Even now, due to the development of the atomic industry, it is possible to produce at atomic stations cheaper energy in kWh cost than at thermal electrical stations. The successful operation of the VVER reactor at the Novo Voronezh Atomic Station and the RBMK reactor at the Leningrad Atomic Station, makes it possible to proceed to the development of more economic thermal reactors with a unit power over 1,500,000 kW. Methods are analysed allowing the atomic industry to be supplied with cheap nuclear fuel on the basis of poor uranium ores. The introduction of radioactive isotopes into the national economy has allowed a number of industries to automate control, to improve technologies and safety measures, etc. Isotopes are being more and more widely used in medicine. Some aspects are considered of using nuclear explosions in the gas and oil industry, in constructing hydraulic engineering works and creating places for the disposal of harmful or radioacmive wastes

Power distribution monitoring and control in 500 MWe PHWR

International Nuclear Information System (INIS)

Kumar, A.

1996-01-01

The 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) is expected to be commissioned in a few years. It has a relatively large sized core with complex material distribution in comparison to the currently operating 220 MWe PHWRs. The resulting neutronically loosely coupled system demands continuous control of the core power distribution. This paper gives a brief description and analysis of the reactor monitoring and control system proposed for this reactor. (author). 11 refs, 8 figs, 3 tabs

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

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Shu-Lin Chen

2016-09-01

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

Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

Science.gov (United States)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-01-01

Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

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

International Nuclear Information System (INIS)

Panus, V.R.

1990-01-01

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

HTR-500 - a technical and engineered safeguards concept

International Nuclear Information System (INIS)

Schoening, J.; Wachholz, W.; Stoelzl, D.

1985-01-01

The plant succeeding the THTR-300 nuclear power plant, which has just started its trial phase of power operation, is the HTR-500. On behalf of the Arbeitsgemeinschaft Hochtemperaturreaktor (AHR), the BBC/HRB Group completed a preliminary project study of a nuclear power plant equipped with a high temperature reactor in the 500 MW power range, in which the changed requirements in the nuclear power market are taken into account and electricity generating costs are to be achieved which are competitive with those of a 1230 MW convoy pressurized water reactor of the present design. On this basis, construction documents are to be drafted, and the licensing procedure under the Atomic Energy Act is to be carried out, within a planning phase of roughly four years. (orig.) [de

Compact Commercial Tokamak Reactor (CCTR): a concept for a 500-MWe commercial-tokamak fusion system

International Nuclear Information System (INIS)

Gillen, T.J.

1980-11-01

A detailed set of self-consistent parameters and costs for the conceptual design of a Compact Commercial Tokamak Reactor (CCTR) is given. Several of the basic design features are the following: an ignited plasma with a major radius of 4.9 m and minor radius of 1.4 m; a net electrical output of 500 MW; a borated-water-cooled, stainless steel shield; and a toroidal field of 12 T at the coil. The design, which utilizes the Westinghouse computer code for the COsting And Sizing of D-T burning Tokamaks (COAST), mainly provides the sizes and geometries associated with the definition of the main component features for which a detailed engineering design can be effectively undertaken. Design study alternatives, including a neutral beam driven design option, a design option with a toroidal field of 13 T at the coil, and a tungsten-shielded option are considered for the CCTR. Also included is the conceptual design of a Compact Fusion Engineering Device

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

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

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

Science.gov (United States)

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

2018-01-13

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

Strategic directions of social management of the Voronezh region economy

Directory of Open Access Journals (Sweden)

A. A. Galkin

2018-01-01

Full Text Available Over the past decades, Russian society has changed significantly. A virtual information and financial space is being formed, where there are no customary boundaries of territories. Social communication begins to develop in new forms, knowledge is exchanged and social consciousness is manipulated. To develop the economy, the most acute social need is its focus on social landmarks. An urgent need is the creation of mechanisms: the approval of uniform, equitable, social bases of legality in all regions of the Russian Federation; maintenance of civil peace; intellectual self–reproduction of the nation, through the appropriate system of education, education, health; reduction of income differentiation; self–realization of man. The ongoing social tolerance is not an expression of the slavish or heroic traits of the nature of Russian citizens, but rather the manifestation of expectations from the political process of what the immature market economy was not capable of accomplishing. Potential factors of the region's strategic development or its competitive advantages are highlighted in the paper. Five strategic directions of social management of the region's economy are grounded: the first is the creation and maintenance of a healthy atmosphere for sustainable development of the region; the second – stimulation of the aggregate consumer demand of the Voronezh region; the third is the democratization of capital; the fourth – the social orientation in agriculture; the fifth – the adjustment of the environmental policy of the region. For each direction, the distribution approaches are highlighted. So, the first direction involves: increasing the social responsibility of entrepreneurs, the entire population, executive and legislative authorities to society. The second is based on the skilful expansion of the intervention of regional and municipal authorities in the economy. The third should involve considerable investments in the regional economy

Kas Gaddafi tõesti põgenes Liibüast? / Heiki Suurkask

Index Scriptorium Estoniae

Suurkask, Heiki, 1972-

2011-01-01

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

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

OpenAIRE

Raab, Andreas

2013-01-01

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

Selection of support structure materials for irradiation experiments in the HFIR [High Flux Isotope Reactor] at temperatures up to 500 degrees C

International Nuclear Information System (INIS)

Farrell, K.; Longest, A.W.

1990-01-01

The key factor in the design of capsules for irradiation of test specimens in the High Flux Isotope Reactor at preselected temperatures up to 500 degree C utilizing nuclear heating is a narrow gas-filled gap which surrounds the specimens and controls the transfer of heat from the specimens through the wall of a containment tube to the reactor cooling water. Maintenance of this gap to close tolerances is dependent on the characteristics of the materials used to support the specimens and isolate them from the water. These support structure materials must have low nuclear heating rates, high thermal conductivities, and good dimensional stabilities under irradiation. These conditions are satisfied by certain aluminum alloys. One of these alloys, a powder metallurgy product containing a fine dispersion of aluminum oxide, is no longer manufactured. A new alloys of this type, with the trade name DISPAL, is determined to be a suitable substitute. 23 refs., 13 figs., 3 tabs

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

CERN Multimedia

Linn, C

2011-01-01

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

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

Directory of Open Access Journals (Sweden)

Hua Xie

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

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

International Nuclear Information System (INIS)

Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

2013-01-01

Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

Review of fast reactor activities in India

International Nuclear Information System (INIS)

Paranjpe, S.R.

1982-01-01

A review of fast reactor activities in India is introduced. One stage of construction of the Fast Breeder Test Reactor (FBTR) and design studies for 500MWe Prototype Fast Breeder Reactor (PFBR) are briefly summarized. The emphasis is on fast reactor physics, materials studies, radiochemistry, and the safety and fuel reprocessing programme

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

Directory of Open Access Journals (Sweden)

Kyota Fukazawa

2014-01-01

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

Design of reactor components (non replaceable) of 500 MWe PHWR for enhanced life

International Nuclear Information System (INIS)

Dwivedi, K.P.; Seth, V.K.

1994-01-01

A nuclear power station is characterised by large initial cost and low operating cost. So a plant which is capable of operating for a longer period of time will be economically more attractive. In the past approach had been to design a nuclear power plant for 30 to 40 years of life time. However, with the improvement in technology and incorporation of redundant and diverse safety features it is now possible to design a nuclear power plant for longer life. Now internationally it is being realised that without sacrificing safety features, plant life should be extended till the cost of maintenance or refurbishment is larger than the cost of the replacement capacity. In order to meet the objective of long life, for the components which cannot be easily replaced the life time of about 100 years is being considered as the design objective. For other items replacement, layout space, shielding, access route and lifting capacity and component design are receiving additional emphasis so as to provide a long total station life time. With the above background, design improvements to enhance the life of reactor components for 500 MWe PHWR namely calandria, end shields and calandria vault liners which cannot be replaced and on which any repair is extremely difficult, have been made. This paper deals with design life of these components and the modifications incorporated in the design. (author). 3 refs., 2 tabs., 3 figs

Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

International Nuclear Information System (INIS)

Bacher, P.

1997-01-01

A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

FBR and RBR particle bed space reactors

International Nuclear Information System (INIS)

Powell, J.R.; Botts, T.E.

1983-01-01

Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10 0 K), high coolant-outlet temperatures (1500 to 3000 0 K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H 2 -cooled mode. The RBR will operate only in the open-cycle H 2 -cooled mode

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

CERN Document Server

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

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

Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

International Nuclear Information System (INIS)

Rastogi, B.P.

1989-01-01

This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

Conceptual design of reactor assembly of prototype fast breeder reactor

International Nuclear Information System (INIS)

Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

1996-01-01

The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

Reliability of fast reactor mixed-oxide fuel during operational transients

International Nuclear Information System (INIS)

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

1991-07-01

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

Nuclear reactor in deep water

International Nuclear Information System (INIS)

Anon.

1980-01-01

Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

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

Science.gov (United States)

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

2015-03-01

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

Physical Characteristics of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy

1994-10-01

The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor

The use of energy analysis and indexes of energy efficiency in nuclear power

International Nuclear Information System (INIS)

D'yakonov, E.I.; Ignatenko, E.I.

1991-01-01

The results of calculating the indexes of energy efficiency for NPPs with the WWER-1000 and RBMK-1000 reactors, heat and power NPPs with the WWER-1000 and dictrict heating NPPs with the AST-500 reactor in three fuel cycles, namely, the open one and with uranium and plutonium recycles, are considered. Complex account for the quantity and quality of produced and consumed energy provides for objective evaluation of the indexes of energy efficiency during comparative analysis of nuclear power plants with different types of reactors. It is shown that complex use of the energy produced at a NPP provides for increase of indexes of energy efficiency. The highest indexes are obtained for heat and power NPP with the WWER-1000 reactor in the open fuel cycle, with uranium and plutonium recycle and for NPP with the WWER-1000 reactor with plutonium recycle

Review of fast reactor activities in India (1982-83)

International Nuclear Information System (INIS)

Paranjpe, S.R.

1983-01-01

A review of fast reactor activities in India in 1982-1983 is given. One stage of construction of Fast Breeder Test Reactor (FBTR) is briefly described. The emphasis is on design studies for the 500 MWe Prototype Fast Breeder Reactor (PFBR). The main features of this design are introduced

A multinode digital control system for 500 MWe PHWR

Energy Technology Data Exchange (ETDEWEB)

Patil, G N; Suresh Babu, R M; Jangra, L R; Das, Shantanu; Mallik, S B [Bhabha Atomic Research Centre, Bombay (India). Reactor Control Div.

1994-12-31

A fault tolerant distributed digital computer system for 500 MWe reactor power regulation is configured around standard microcomputer boards designed indigenously. The system is configured as functionally partitioned distributed control system having 8 nodes linked by high-speed dual redundant high-way. The paper gives the details of the configuration of system and how the features of fault-tolerance and fail-safeness are achieved through design. (author). 1 fig.

Single Point Vulnerability Analysis of Automatic Seismic Trip System

International Nuclear Information System (INIS)

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

2016-01-01

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

Single Point Vulnerability Analysis of Automatic Seismic Trip System

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

Science.gov (United States)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

CERN Document Server

The ATLAS collaboration

2016-01-01

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

The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

Science.gov (United States)

Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

1991-02-01

The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

Overview of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

2016-01-01

The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-01

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

The HTR 500 concept based on pratical THTR and AVR experience

International Nuclear Information System (INIS)

Wachholz, W.; Weicht, U.

1988-01-01

This paper discusses progress during the past ten years in the development of a specific HTR safety concept. This has been mainly characterized by the abandonment of the LWR specific safety principles and making use of the safety characteristics typical of the high-temperature reactor (HTR). In the design, construction and operation of high-temperature reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - experience has been gained in the field of accident topology and plant risk of HTRs in recent years. This experience, based on detailed accident analyses performed by manufacturers and experts, is relevant for the entire HTR line independent of specific projects. The authors focus on the HTR 500, the first commercial high temperature reactor with a pebble bed core. Its design principles and the design of its systems are based on the earlier AVR and THTR projects

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

Energy Technology Data Exchange (ETDEWEB)

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

1982-02-12

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

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

Directory of Open Access Journals (Sweden)

Juliana V.F. Sales

2013-01-01

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

Nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-06-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

Status of development of gas-cooled reactors in Switzerland, 1987

International Nuclear Information System (INIS)

Helbling, W.; Sarlos, G.

1988-01-01

Swiss industrial companies and the Federal Institute for Reactor Research are involved in the framework of German-Swiss cooperation in the HTR-500 project. Another effort in Switzerland is directed to the development of a small heating reactor, in the power range of 10 to 50 MW, for district heating, one of the concepts investigated being an HTR pebble-bed reactor

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

International Nuclear Information System (INIS)

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

2011-01-01

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

AzTEC on ASTE Survey of Submillimeter Galaxies

Science.gov (United States)

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

2008-10-01

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

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

Science.gov (United States)

2011-08-18

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

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

Development of manufacturing process for production of 500 MWe calandria sheets

International Nuclear Information System (INIS)

Hariharan, R.; Ramesh, P.; Lakshminarayana, B.; Bhaskara Rao, C.V.; Pande, P.; Agarwala, G.C.

1992-01-01

Calandria tubes made of zircaloy-2 are being used as structural components in pressurised heavy water power reactors. The sheets required for producing calandria tube for 235 MWe reactors are being manufactured at Zircaloy Fabrication Plant (ZFP), NFC utilizing a 2 Hi/4 Hi rolling mill procured for the purpose, by carrying out cold rolling process to achieve the required size after hot rolling suitable extruded slabs. Due to limitation of width of the sheet that can be rolled with the mill as well as the size of the slab that can be extruded with the existing press, difficulties arose in producing acceptable full length sheets of size 6600 mm long x 435 mm wide x 1.6 mm thick for manufacturing 500 MWe calandria tube. This paper deals with the details of the process problem resolved. They are: (a)designing of suitable hot and cold rolling pass schedules, (b)selection and standardization of process parameters such as beta quenching, hot rolling and cold rolling, and (c)details of the overall manufacturing process. Due to implementation of above, sheets required for manufacturing 500 MWe calandria tube sheets were successfully rolled. About 40 nos. of acceptable full length sheets have already been manufactured. (author). 1 fig., 3 tabs

Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

International Nuclear Information System (INIS)

Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

1981-01-01

Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

Comparison of Pickering NGS performance with world power reactors, 1977

International Nuclear Information System (INIS)

Buhay, S.

Pickering NGS performance is compared, in highly graphic form, with the perfomance of other nuclear power plants around the world. The four Pickering reactors score in the top six, rated by gross capacity factor. Major system suppliers for world power reactors above 500 MW are cataloged. (E.C.B.)

REGIONAL COMPONENTS OF GROWTH THE QUALITY OF LIFE OF THE POPULATION (FOR EXAMPLE, VORONEZH REGION

Directory of Open Access Journals (Sweden)

N. A. Serebryakova

2015-01-01

Full Text Available At the present stage of development of economy there is a transformation of main objectives of the state from ensuring growth of welfare of citizens to ensuring continuous growth of quality of life. Ensuring growth of quality of life of the population is carried out not only on state, but also on regional levels. At the state level key tasks are formed, the state assumes a considerable share of functions on achievement of the designated purpose. In the real research the attention to consideration of the program documents directed on improvement of quality of life of the population on the example of the Voronezh region is paid. The program documents of a social orientation existing in the region are for this purpose studied. It is revealed that in the region a number of the programs and subprogrammes aimed at improvement of a demographic situation, improvement of a regional budgetary and tax policy, social support of citizens in general works. As tools which use regional authorities, are noted: social standards, address social help, social contracts. It is established that introduction of system of social contracts allows to save budget funds and, at the same time, to motivate the persons which are in a difficult life situation on change of the situation to the best, joint efforts with social security authorities. It is noted that as the leading principles of rendering the social help paramount value has detailed definition of degree of need of the help and a condition of granting. The last often generate dependant moods which are important for leveling, using levers of social interaction. The analysis of the contents of the realized programs for improvement of quality of life of the population allowed to reveal advantages and defects in these programs, and also to define reserves of growth of quality of life. So, it is specified that program documents of regional level in the prevailing majority have a narrow focus on ensuring the help and

Development of a model and test equipment for cold flow tests at 500 atm of small nuclear light bulb configurations

Science.gov (United States)

Jaminet, J. F.

1972-01-01

A model and test equipment were developed and cold-flow-tested at greater than 500 atm in preparation for future high-pressure rf plasma experiments and in-reactor tests with small nuclear light bulb configurations. With minor exceptions, the model chamber is similar in design and dimensions to a proposed in-reactor geometry for tests with fissioning uranium plasmas in the nuclear furnace. The model and the equipment were designed for use with the UARL 1.2-MW rf induction heater in tests with rf plasmas at pressures up to 500 atm. A series of cold-flow tests of the model was then conducted at pressures up to about 510 atm. At 504 atm, the flow rates of argon and cooling water were 3.35 liter/sec (STP) and 26 gal/min, respectively. It was demonstrated that the model is capable of being operated for extended periods at the 500-atm pressure level and is, therefore, ready for use in initial high-pressure rf plasma experiments.

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

Science.gov (United States)

Megías, Eugenio; Quirós, Mariano; Salas, Lindber

2017-07-01

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

Qualification of PHT piping of Indian 500 MW PHWR for LBB, using R-6 method

International Nuclear Information System (INIS)

Rastogi, Rohit; Bhasin, V.; Kushwaha, H.S.

1997-01-01

This document discusses the qualification of straight pipe portion of the primary heat transport (PHT) piping of Indian 500 MWe pressurised heavy water reactor (PHWR) for leak before break (LBB). The evaluation is done using R-6 [1] method. The results presented here are: the safety margins which exist on straight pipe components of main PHT piping of 500 MWe, under leakage size crack (LSC) and design basis accident loads; the sensitivity of safety margins with respect to different analysis parameters and the qualification of PHT piping for LBB based on criterion given by NUREG-1061 [2] and TECDOC-774 [3]. (author)

E-beam heated linear solenoid reactors

International Nuclear Information System (INIS)

Benford, J.; Bailey, V.; Oliver, D.

1976-01-01

A conceptual design and system analysis shows that electron beam heated linear solenoidal reactors are attractive for near term applications which can use low gain fusion sources. Complete plant designs have been generated for fusion based breeders of fissile fuel over a wide range of component parameters (e.g., magnetic fields, reactor lengths, plasma densities) and design options (e.g., various radial and axial loss mechanisms). It appears possible that a reactor of 100 to 300 meters length operating at power levels of 1000 MWt can economically produce 2000 to 8000 kg/yr of 233 U to supply light water reactor fuel needs beyond 2000 A.D. Pure fusion reactors of 300 to 500 meter lengths are possible. Physics and operational features of reactors are described. Beam heating by classical and anomalous energy deposition is reviewed. The technology of the required beams has been developed to MJ/pulse levels, within a factor of 20 of that needed for full scale production reactors. The required repetitive pulsing appears practical

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

Science.gov (United States)

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

2017-12-01

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

Compatibility of vanadium alloys with reactor-grade helium for fusion reactor applications

International Nuclear Information System (INIS)

Bell, G.E.C.; Bishop, P.S.

1993-01-01

Tests were conducted to determine the compatibility of vanadium alloys with reactor-grade helium and to define the helium gas chemistry requirements for fusion reactors, miniature tensile specimens of V-5Cr-5Ti. V-10Cr-5Ti, and V-12.5Cr-5 Ti were exposed in a once-through system to helium with 70 vppm-H 2 (measured oxygen partial pressures of 10 -12 atm) and bottle helium (measured oxygen partial pressures of -4 atm) between 500 and 700 degree C for up to 1008 h. The weight changes in the specimens were recorded. The helium-exposed specimens were tensile tested, and the effects of exposure on mechanical properties were assessed. Exposure between 500 and 700 degree C for 1008 h in He+70 vppm-H 2 resulted in complete embrittlement of all the alloys in room temperature tensile tests. The fracture mode was primarily cleavage, probably caused by a hydrogen-induced shift in the ductile to brittle transition temperature (DBTT). Weight gains increased with temperature and were largest for the V-5Cr-5Ti alloy. Specimens exposed for 531 h between 500 and 700 degree C in bottle He exhibited two distinct fracture morphologies on the fracture surfaces. Brittle cleavage around the edges of specimens gave way to ductile dimpling in the center of the specimens. The brittle region around the periphery of the specimen is most likely the highest vanadium oxide. V 2 O 5

Role of pressuriser in enhancing pressure control system capability in primary system of 500 MWe PHWR

Energy Technology Data Exchange (ETDEWEB)

Walia, M P.S.; Misri, Vijay; Bapat, C N; Sharma, V K [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

The primary heat transport system of a pressurized heavy water reactor (PHWR) extracts and transports the heat produced in the fuel (located inside coolant channel assemblies) to the steam generators where steam is generated to run the turbo-generator. The heat transport medium (primary coolant) is heavy water which is kept in a pressurized liquid state with the help of a pressure control system. Feed and bleed circuits with associated equipment of PHT main system have traditionally constituted the pressure control system. However, for large size reactors of 500 MWe capacity, a surge tank known as pressurizer was incorporated due to the presence of relatively large inventory in PHT main circuit. The pressurizer acts as a cushion for pressure variations resulting from various transients. This significantly reduces the onerous demand on feed and bleed system, thereby reducing reactor outages on system pressure excursions. The paper describes in detail the pressure control system of 500 MWe PHWR involving pressuriser and feed and bleed system including their functions and instrumentation. The results of mathematical modelling/analysis undertaken to establish the response adequacy of pressure control system, to postulated plant transients vis-a-vis the role of pressurizer are presented. (author). 10 figs.

Present status of space nuclear reactor

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

USA and former USSR led space development, and had the experience of launching nuclear reactor satellites. In USA, the research and development of space nuclear reactor were advanced mainly by NASA, and in 1965, the nuclear reactor for power source ''SNAP-10A'' was launched and put on the orbit around the earth. Thereafter, the reactor was started up, and the verifying test at 500 We was successfully carried out. Also for developing the reactor for thermal propulsion, NERVA/ROVER project was done till 1973, and the technological basis was established. The space Exploration Initiative for sending mankind to other solar system planets than the earth is the essential point of the future projects. In former USSR, the ground experiment of the reactor for 800 We power source ''Romashka'', the development of the reactor for 10 kWe power source ''Topaz-1 and 2'', the flight of the artificial satellites, Cosmos 954 and Cosmos 1900, on which nuclear reactors were mounted, and the operation of 33 ocean-monitoring satellites ''RORSAT'' using small fast reactors were carried out. The mission of space development and the nuclear reactors as power source, the engineering of space nuclear reactors, the present status and the trend of space nuclear reactor development, and the investigation by the UN working group on the safety problem of space nuclear reactors are described. (K.I.)

Truth-telling contra perfectionist liberalism: Muslim parrhēsíastes in Denmark

DEFF Research Database (Denmark)

Renders, Johannes

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

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

2003-01-01

Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

Liberalism as a Cultural Phenomenon of Russian Provincial Life of the Late Nineteenth – Early Twentieth Centuries (according to Voronezh province

Directory of Open Access Journals (Sweden)

Evgeniy V. Dvoretskiy

2016-12-01

Full Text Available The authors of the presented research considers liberalism as a cultural phenomenon, closely watches the liberalism as an impact factor to cultural life in rural areas (esp. in Voronezh province in the end of XIX up to XX cent.. Besides, the authors make a try of consideration of liberalism in the context of European innovations blending conception. According to this idea the traditional society was transformed by means of technical, political, social and cultural innovation expansion from their source, which is in the Western Europe. Different sources served as a basis for the research, their analysis gives the reason to shape out the place and the role of Russian liberalism in the end of XIX up to the beginning of XX cent. in the formation of innovative elements of urban culture, and to specify the process of the liberal ideas expansion in the provincial cultural life sphere, and to characterize the bearer and translators of liberal ideas (local liberal figures, participants of the local liberal parties, liberal intellectuals. As a result, the existence of the basic liberal values reflections (freedom and preciousness of a personal identity, individualism and so on was stated, that shows the Russian liberalism as the ideal, political and cultural phenomenon. This, finally, states the great degree of the reception of the liberal ideas by the Russian society in the beginning of the XXth, and gives the opportunity to observe the liberal reflections as a process of civil society’s formation evidence in Russia by the beginning of the XXth cent.

Mechanical design of a PERMCAT reactor module

Energy Technology Data Exchange (ETDEWEB)

Tosti, S. [Associazione ENEA Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy)], E-mail: tosti@frascati.enea.it; Bettinali, L. [Associazione ENEA Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Borgognoni, F. [Tesi Sas, Via Bolzano 28, Rome (Italy); Murdoch, D.K. [EFDA CSU, Boltzmannstr. 2, D-85748 Garching bei Munchen (Germany)

2007-02-15

The PERMCAT is a membrane reactor proposed for processing fusion reactor plasma exhaust gas: tritium removal is obtained by isotopic swamping operating in counter-current mode. In this work, a membrane reactor using a permeator tube of length about 500 mm produced via diffusion welding of Pd-Ag thin foils is described. An appropriate mechanical design of the membrane module has been developed in order to avoid any significant compressive and bending stresses on the very long and thin wall permeator tube: two expanded bellows have been applied to the Pd-Ag tube, so that it has been pre-tensioned before operating. The elongation of the metal permeator under hydrogenation has been theoretically estimated and experimentally verified for properly designing the membrane reactor.

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

OpenAIRE

Trugeon, Mélanie

2011-01-01

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

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

Directory of Open Access Journals (Sweden)

Ya-Ling Wang

2016-09-01

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

The control of emissions from nuclear power reactors in Canada

International Nuclear Information System (INIS)

Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

1988-01-01

Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

Laser fusion hybrid reactor systems study

International Nuclear Information System (INIS)

1976-07-01

The work was performed in three phases. The first phase included a review of the many possible laser-reactor-blanket combinations and resulted in the selection of a ''demonstration size'' 500 MWe plant for further study. A number of fast fission blankets using uranium metal, uranium-molybdenum alloy, and uranium carbide as fuel were investigated. The second phase included design of the reactor vessel and internals, heat transfer system, tritium processing system, and the balance of plant, excluding the laser building and equipment. A fuel management scheme was developed, safety considerations were reviewed, and capital and operating costs were estimated. Costs developed during the second phase were unexpectedly high, and a thorough review indicated considerable unit cost savings could be obtained by scaling the plant to a larger size. Accordingly, a third phase was added to the original scope, encompassing the redesign and scaling of the plant from 500 MWe to 1200 MWe

Instrumentation and control for reactor power setback in PFBR

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Vasal, Tanmay; Nagaraj, C.P.; Madhusoodanan, K.

2013-01-01

In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

Comments on nuclear reactor safety in Ontario

International Nuclear Information System (INIS)

1987-08-01

The Chalk River Technicians and Technologists Union representing 500 technical employees at the Chalk River Nuclear Laboratories of AECL submit comments on nuclear reactor safety to the Ontario Nuclear Safety Review. Issues identified by the Review Commissioner are addressed from the perspective of both a labour organization and experience in the nuclear R and D field. In general, Local 1568 believes Ontario's CANDU nuclear reactors are not only safe but also essential to the continued economic prosperity of the province

A simplified model of aerosol removal by natural processes in reactor containments

Energy Technology Data Exchange (ETDEWEB)

Powers, D.A.; Washington, K.E.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States); Burson, S.B. [Nuclear Regulatory Commission, Washington, DC (United States)

1996-07-01

Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.

Protection of semiconductor converters for controlled bypass reactors

International Nuclear Information System (INIS)

Dolgopolov, A. G.; Akhmetzhanov, N. G.; Karmanov, V. F.

2010-01-01

Possible ways of protecting thyristor converters in systems for magnetizing 110 - 500 kV controlled bypass reactors during switching and automatic reclosing are examined based on experience with the development of equipment, line tests, and mathematical modelling.

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

Science.gov (United States)

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

2008-08-01

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

Prevention of biofouling and biocorrosion in reactor systems

International Nuclear Information System (INIS)

Mathur, A.K.; Shivananda, S.R.

1995-01-01

Formaldehyde even at 500 μl/l concentration can prevent the growth of bacteria, algae and fungi in thermal reactors there by stopping the chances of biocorrosion and plugging of the pipe lines. (author). 5 refs., 1 fig

The status of development of small and medium sized reactors

International Nuclear Information System (INIS)

Konstantinov, L.V; Kupitz, J.

1987-01-01

Several IAEA Member States have shown their interest in reactor design, having a smaller power rating (100-500 MW(e) range) than those generally available on the international market. These small and medium sized power reactors are of interest either for domestic applications or for export into countries with less developed infrastructure. There are different developments undertaken for these power reactors to be ready for offering in the nineties and beyond. The paper gives an overview about the status and different trends in IAEA Member States in the development of small and medium sized reactors for the 90's and provides an outlook for very new reactor designs as a long term option for nuclear power. (author)

Related activities on management of ageing of Dalat Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lam, Pham Van [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

1998-10-01

The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

Walter, C.E.; Pearson, J.S.

1987-05-01

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

Designing a nuclear power plant with 1000 MW WWER-type units

Energy Technology Data Exchange (ETDEWEB)

Berkovich, V; Kaloshin, J; Tatarnikov, V; Shenderovich, A

1977-06-01

A brief description is presented of a WWER-1000 nuclear power plant also considering its environmental impact and the problem of core poisoning. The following indicators are graphically shown in relation to the reactor output: turbogenerator unit outputs, efficiency, specific capital costs and own costs of electric power generated by the Voronezh nuclear power plant. Also listed are the specific consumption of metal and concrete, specific equipment weight and the specific volume of the buildings of the main generating unit as well as the cross section thereof.

Preliminary safety analysis of molten salt breeder reactor

International Nuclear Information System (INIS)

Cheng Maosong; Dai Zhimin

2013-01-01

Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)

Current status of operation and utilization of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Le Van So

2004-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW swimming pool type reactor using the Soviet WWR-SM fuel assembly with 36% enrichment of U-235. It was upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analyses and research purposes. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for physics experiments and training purpose. From the first start-up to the end of December 2002, it totaled about 24,700 hrs of operation and the total energy released was 490 MWd. After 10 years of operation with the core of 89-fuel assembly configuration, in April 1994, the first refueling work was done and the 100-fuel assembly configuration was set-up. The second fuel reloading was executed in March 2002. At present time, the working configuration of the reactor core consists of 104 fuel assemblies. This fuel reloading will ensure efficient exploitation of the reactor for about 3 years with 1200-1300 hrs per year at nominal power. The current status of operation and utilization and some activities related to the reactor core management of the DNRR are presented and discussed in this paper. (author)

Liquid metal reactor core material HT9

International Nuclear Information System (INIS)

Kim, S. H.; Kuk, I. H.; Ryu, W. S. and others

1998-03-01

A state-of-the art is surveyed on the liquid metal reactor core materials HT9. The purpose of this report is to give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is planned for the year of 2010. In-core stability of cladding materials is important to the extension of fuel burnup. Austenitic stainless steel (AISI 316) has been used as core material in the early LMR due to the good mechanical properties at high temperatures, but it has been found to show a poor swelling resistance. So many efforts have been made to solve this problem that HT9 have been developed. HT9 is 12Cr-1MoVW steel. The microstructure of HT9 consisted of tempered martensite with dispersed carbide. HT9 has superior irradiation swelling resistance as other BCC metals, and good sodium compatibility. HT9 has also a good irradiation creep properties below 500 dg C, but irradiation creep properties are degraded above 500 dg C. Researches are currently in progress to modify the HT9 in order to improve the irradiation creep properties above 500 dg C. New design studies for decreasing the core temperature below 500 dg C are needed to use HT9 as a core material. On the contrary, decrease of the thermal efficiency may occur due to lower-down of the operation temperature. (author). 51 refs., 6 tabs., 19 figs

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

Science.gov (United States)

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

2014-06-19

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

Status of Dalat research reactor and progress of new reactor plan in Vietnam

International Nuclear Information System (INIS)

Dien, Nguyen Nhi; Vien, Luong Ba

2005-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500-kW pool-type reactor loaded with the Soviet WWR-M2 Fuel Assemblies (FA), moderated and cooled by light water. The reactor was reconstructed from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The first criticality of the renovated reactor was achieved on 1 st November 1983, and then on 20 March 1984 the reactor was officially inaugurated and its activities restarted. During the last twenty years, the DNRR has played an important role as a large national research facility to implement researches and applications, and its utilization has been broadened in various fields of human life. However, due to the limitation of the neutron flux and power level, the out-of date design of the experimental facilities and the ageing of the reactor facilities, it cannot meet the increasing user's demands even in the existing utilization areas. In addition, the utilization demands of the Research Reactor (RR) will be increased along with the development of the nation's economy growth. In this aspect, it is necessary to have in Vietnam a new high performance multipurpose RR with a sufficient neutron flux and power level. According to the last draft of a national strategy for atomic energy development submitted to the Government for consideration and approval, it is expected that a new high power RR would be put into operation before 2020. The operation and utilization status of the DNRR is presented and some preliminary results of the national research project on new reactor plan for Vietnam are discussed in this paper

Review of current and proposed reactor upgrades

International Nuclear Information System (INIS)

Moon, R.M.

1985-01-01

In an effort to foresee the future health of neutron scattering, a survey of plans to upgrade reactors and associated experimental facilities was undertaken. The results indicate that we are now entering a period characterized by a substantial reinvestment in reactor sources and expansion in the number of neutron scattering instruments. For the group of institutions participating in this survey there will be a total investment in improved sources and experimental facilities of $500 M to $1,000 M over the next decade. This investment will result in a 30 to 40% increase in the total power of research reactors and an increase of 30 to 50% in the number of neutron scattering instruments. It is therefore reasonable to anticipate an approximate doubling in the number of reactor neutrons incident on samples in the mid 90s compared to the present

Power start up of the Dalat nuclear research reactor

International Nuclear Information System (INIS)

Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

1994-01-01

After accomplishing the physical start-up of the reactor, the power start-up was carried out in February 1984. The power of the reactor has reached: 10 KW on 6/2/1984, 100 KW on 7/2/1984, 200 KW and 300 KW on 8/2/1984; 400 KW and nominal power 500 KW on 9/2/1984. The reactivity temperature coefficient and the xenon poisoning were determined. 3 figs., 12 tabs

Natural Circulation Capability Assessments for a Small-medium Reactor

International Nuclear Information System (INIS)

Choi, Sun Do

2010-02-01

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

Some corrosion effects of the aluminum tank surface of Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Mong Sinh

1995-01-01

The Dalat Nuclear Research Reactor was reconstructed from the TRIGA-MARK-II reactor installed in 1963 with a nominal power of 250 kW. Reconstruction and upgrading of this reactor to nominal power of 500 kW had been completed in the end of 1983. The reactor was commissioned in the beginning of March 1984. The aluminum reactor tank and some components of the former reactor are more than 30 year old. The good quality of reactor water minimized the total corrosion rate of reactor material surface. But some local corrosion had been found out at the tank bottom especially in water stagnant areas. The corrosion processes could be due to the electrochemical reactions associated with different metals and alloys in the reactor water and keeping in touch with the surface of aluminum reactor tank. (orig.)

Design and analysis on super-critical water cooled power reactors

International Nuclear Information System (INIS)

Ishiwatari, Yuki

2005-01-01

The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

Evolution of MMI for 500 MWe PHWR plant

International Nuclear Information System (INIS)

Surendar, Ch.; Sharma, M.P.; Jayanthi, S.

1994-01-01

The Indian nuclear power programme for building Pressurized Heavy Water Reactors began with the construction of two units at Kota, Rajasthan. Although the concept of a centralized control room has been used since the beginning, the man-machine interface design has evolved with technological developments. The man-machine interaction in the earliest plants imposed a considerable burden on the operators and led to a need for more sophisticated instrumentation. Several microprocessor and computer based systems were identified and developed and many were retrofitted into existing plants providing immediate advantages. This paper traces the evolution of many of these systems and also describes the basis and the architecture for the man-machine interaction scheme in the 500 MWe nuclear power plants currently being designed. (author). 7 refs., 2 figs., 1 tab

Core design concepts for high performance light water reactors

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.

2007-01-01

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

Designing a nuclear power plant with 1000 MW WWER-type units

International Nuclear Information System (INIS)

Berkovich, V.; Kaloshin, J.; Tatarnikov, V.; Shenderovich, A.

1977-01-01

A brief description is presented of a WWER-1000 nuclear power plant also considering its environmental impact and the problem of core poisoning. The following indicators are graphically shown in relation to the reactor output: turbogenerator unit outputs, efficiency, specific capital costs and own costs of electric power generated by the Voronezh nuclear power plant. Also listed are the specific consumption of metal and concrete, specific equipment weight and the specific volume of the buildings of the main generating unit as well as the cross section thereof. (J.B.)

Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

Energy Technology Data Exchange (ETDEWEB)

Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

2013-01-29

Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

Status report on nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-01-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

Current status of operation and utilization of the Dalat Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien

2006-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool-type reactor using the Soviet VVR-M2 fuel assembly with 36% enrichment of U-235. It was renovated and upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analysis, scientific research and training. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for reactor physics and thermal hydraulics experiments. From the first start-up to the end of December 2003, it totaled about 26,000 hrs of operation and the total energy released was about 515 MWd. After 10 years of operation with the core of 89-fuel assembly configuration, in April 1994, the first refueling work was done and the 100-fuel assembly configuration was set-up. The second fuel reloading was executed in March 2002. At present time, the working configuration of the reactor core consists of 104 fuel assemblies. The next fuel reloading has been planned at the end of 2004. The current status of operation and utilization of the DNRR is presented and discussed in this paper. (author)

Comparison of Ontario Hydro's performance with world power reactors - 1981

International Nuclear Information System (INIS)

Dumka, B.R.

1982-04-01

The performance of Ontario Hydro's CANDU reactors in 1981 is compared with that of 123 world nuclear power reactors rated at 500 MW(e) or greater. The report is based on data extracted from publications, as well as correspondence with a number of utilities. The basis used is the gross capacity factor, which is defined as gross unit generation divided by the perfect gross output for the period of interest. The lowest of the published turbine and generator design ratings is used to determine the perfect gross output, unless the unit has been proven capable of consistently exceeding this value. The first six reactors in the rankings were CANDU reactors operated by Ontario Hydro

Review of fast reactor activities in India (1984)

International Nuclear Information System (INIS)

Paranjpe, S.R.

1986-01-01

During the year a number of reviews and construction activities have been practically completed as required for the 1st criticality of FBTR. The reactor is expected to become critical by the middle of 1985. The design studies for 500 MWe prototype fast breeder reactor (PFBR) have been continued. Due to preoccupation with the completion of construction of FBTR, the limited effort has been focussed on the design of key components like the sodium pumps, drivers for sodium pumps, control rod drive mechanism and steam generators. The main programs, which are a continuing activity in RRC, are discussed in this report. They are: reactor physics, radio-chemistry, metallurgy, reprocessing and safety research

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

CERN Document Server

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

2016-01-01

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

Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

International Nuclear Information System (INIS)

Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

2003-01-01

Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

Commercial Aspect of Research Reactor Fuel Element Production

International Nuclear Information System (INIS)

Susanto, B.G; Suripto, A

1998-01-01

Several aspects affecting the commercialization of the Research Reactor Fuel Element Production Installation (RR FEPI) under a BUMN (state-owned company)have been studied. The break event point (BEP) value based on total production cost used is greatly depending upon the unit selling price of the fuel element. At a selling price of USD 43,500/fuel element, the results of analysis shows that the BEP will be reached at 51% of minimum available capacity. At a selling price of US$ 43.500/fuel element the total income (after tax) for 7 years ahead is US $ 4.620.191,- The net present value in this study has a positive value is equal to US $ 2.827.527,- the internal rate of return will be 18% which is higher than normal the bank interest rare (in US dollar) at this time. It is concluded therefore that the nuclear research reactor fuel element produced by state-owned company BUMN has a good prospect to be sold commercially

Compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

Startup testing of Romania dual-core test reactor

International Nuclear Information System (INIS)

Whittemore, W.L.

1980-01-01

Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

Design study of 'HIBLIC-I' reactor cavity

International Nuclear Information System (INIS)

Fujiie, Y.

1984-01-01

A preliminary conceptual design of a reactor cavity for HIBLIC-1, a heavy ion fusion reactor system, was carried out. Design efforts have been concentrated mainly on the feasibility study of the physical scenario adopted and also on the system integration of the structures and components into a compact reactor cavity. The design features of the reactor are a compact reactor cavity, maximum coolant temperature up to 500 deg C, the protection of the sacrificial wall and cavity wall from radiation, the protection of the sacrificial wall from the pressure transient due to rapid heating, the selection of a ferritic steel HT-9 as the structural material and impurity control, and tritium breeding and recovery. The purpose of this paper is to describe the outline of the reactor cavity design of HIBLIC-1. The objectives of the preliminary conceptual design were to propose the idea and concept in order to constitute the physical scenario without contradiction and to find out the critical and fundamental problems to be studied in future. The cavity configuration and dynamics, tritium breeding and radiation damage, the behavior of a structural material in liquid lithium and tritium recovery are reported. (Kako, I.)

Small reactors and the 'second nuclear era'

International Nuclear Information System (INIS)

Egan, J.R.

1984-01-01

Predictions of the nuclear industry's demise are premature and distort both history and politics. The industry is reemerging in a form commensurate with the priorities of those people and nations controlling the global forces of production. The current lull in plant orders is due primarily to the world recession and to factors related specifically to reactor size. Traditional economies of scale for nuclear plants have been greatly exaggerated. Reactor vendors and governments in Great Britain, France, West Germany, Japan, the United States, Sweden, Canada, and the Soviet Union are developing small reactors for both domestic applications and export to the Third World. The prefabricated, factory-assembled plants under 500 MWe may alleviate many of the existing socioeconomic constraints on nuclear manufacturing, construction, and operation. In the industrialized world, small reactors could furnish a qualitatively new energy option for utilities. But developing nations hold the largest potential market for small reactors due to the modest size of their electrical systems. These units could double or triple the market potential for nuclear power in this century. Small reactors will both qualitatively and quantitatively change the nature of nuclear technology transfers, offering unique advantages and problems vis-a-vis conventional arrangements. (author)

Use of TCSR with Split Windings for Shortening the Spar Cycle Time in 500 kV Lines

Energy Technology Data Exchange (ETDEWEB)

Matinyan, A. M., E-mail: al-drm@mail.ru; Peshkov, M. V.; Karpov, V. N.; Alekseev, N. A. [JSC “R& D Center at Federal Grid Company of Unified Power System,” (Russian Federation)

2017-01-15

The arc-fault recharge phenomenon in single-phase automatic reclosure (SPAR) of a line is examined. Abrief description is given of the design of a 500 kV thyristor controlled shunt reactor (TCSR) with split valve-side windings. This type of TCSR is shown to effectively quench a single-phase arc fault in a power transmission line and shortens the SPAR cycle time.

Heavy water cycle in the CANDU reactor

International Nuclear Information System (INIS)

Nanis, R.

2000-01-01

Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

Nuclear Reactor Sharing Program

International Nuclear Information System (INIS)

1994-01-01

The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

International Nuclear Information System (INIS)

Dhami, P.S; Yadav, J.S; Agarwal, K.

2017-01-01

Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

Use of research reactors in multidisciplinary education at Cornell University

International Nuclear Information System (INIS)

Clark, D.D.

1992-01-01

Multidisciplinary aspects of nuclear science and technology form a large part of the research and teaching activities of the Nuclear Science and Engineering (NS and E) Program at Cornell, and the two reactors housed in Ward Laboratory - a 500-kW TRIGA and a 100-W critical facility [zero-power reactor (ZPR)]- play a central role in those activities. Several primarily educational and multidisciplinary features of the NS and E program are described in this paper

Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

International Nuclear Information System (INIS)

Singh, Ashutosh Pratap; Rajagopalan, C.; Rakesh, V.; Rajendran, S.; Venugopal, S.; Kasiviswanathan, K.V.; Jayakumar, T.

2011-01-01

Highlights: → Conceptual study on the configuration of an ISI device for FBR interspace environment has been carried out. → Prototyping of the concept has been experimentally validated in a mock up. → High temperature version of the ISI device has been made and tested in mock-up. Further experimentation is underway. → Simulation of different configurations of the device has been carried out with respect to reduced gap between main vessel and safety vessel for future FBRs. → Studies on wheel lining for the device have been carried out at 150 o C for better traction and payload capability. - Abstract: In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.

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

Contributions of fast breeder test reactor to the advanced technology in India

International Nuclear Information System (INIS)

Kapoor, R.P.

2001-01-01

Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe loop type, sodium cooled, plutonium rich mixed carbide fuelled reactor. Its operation at Indira Gandhi Centre for Atomic Research, since first criticality in 1985, has contributed immensely to the advancement of this multidisciplinary and complex fast breeder technology in the country. It has also given a valuable operational feedback for the design of 500 MWe Prototype Fast Breeder Reactor. This paper highlights FBTR's significant contributions to this important technology which has a potential to provide energy security to the country in future. (author)

Characteristic thermal-hydraulic problems in NHRs: Overview of experimental investigations and computer codes

Energy Technology Data Exchange (ETDEWEB)

Falikov, A A; Vakhrushev, V V; Kuul, V S; Samoilov, O B; Tarasov, G I [OKBM, Nizhny Novgorod (Russian Federation)

1997-09-01

The paper briefly reviews the specific thermal-hydraulic problems for AST-type NHRs, the experimental investigations that have been carried out in the RF, and the design procedures and computer codes used for AST-500 thermohydraulic characteristics and safety validation. (author). 13 refs, 10 figs, 1 tab.

A review of the Indian fast reactor programme

International Nuclear Information System (INIS)

Bhoje, S.B.

1990-01-01

Development of Fast Breeder activities is being done mainly at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam and the total Scientific and Technical staff working at the Centre for development of FBRs is about 1200. The development work relating to the fuel fabrication and design and development for some of the fuel handling equipment is being done at the Bhabha Atomic Research Centre, Trombay, Bombay. Complete recovery from the fuel handling incident of FBTR was achieved during the beginning of 1989. Damaged guide tube and bent subassemblies were replaced, the incident was analysed in detail and appropriate remedial measures, viz., modifications in the fuel handling machine control logic and plug rotation logic were implemented to prevent its recurrence. Safety clearances for the restart of the reactor were obtained from the Atomic Energy Regulatory Board in May 1989. As steam generators were not valved in the secondary sodium system, the reactor power during this phase of operation was limited to 500 KWt. The main objectives during this phase were to complete the balance low power physics experiments and to operate the reactor for a sufficiently long time to assess the performance of various systems, in particular the neutronic instrumentation, control rod drive and safety logic system which were not in active service for the two years. From May to July, 1989, the reactor was successfully operated up to a power level of 500 KWt with 50% operating time. Design of PFBR is progressing intensively. (author). 1 tab

Development of process route for production of tubing for various core sub-assemblies and heat exchangers for 500 MWe Indian PFBR

International Nuclear Information System (INIS)

Lakshminarayana, B.; Phani Babu, C.; Dubey, A.K.; Surender, A.; Deshpande, K.V.K.; Maity, P.K.

2009-01-01

India's three stage Nuclear Power Program has entered its second stage on commercial scale with the commencement of construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. Nuclear Fuel Complex (NFC), Hyderabad is playing a crucial role in the manufacture of all the critical sub-assemblies and control elements for this reactor. The challenging task of process development and production of the various critical tubing for these sub assemblies for PFBR has been taken up by Stainless Steel Tubes Plant (SSTP), NFC with indigenous development of the equipment and technology

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

Science.gov (United States)

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

2010-11-01

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

Activity report of Reactor Physics Division : 1990

International Nuclear Information System (INIS)

Mohanakrishnan, P.

1991-01-01

The major Research and Development and Project activities carried out during the year 1990 in Reactor Physics Division are presented in the form of summaries in this report. The various activities are organised under the following areas : (1) Nuclear Data Evaluation, Processing and Validation, (2) Core Physics and Analysis, (3) Reactor Kinetics and Safety Analysis, (4) Noise Analysis, and (5) Radiation Transport and Shielding. FBTR was restarted in July 1990 and the power was raised upto 500 kW. A number of low power physics experiments on reactivity coefficients, kinetics and noise, neutron flux and gamma dose in B cells, were performed, which are discussed in this report. (author). figs., tabs

Prototype fast breeder reactor main options

International Nuclear Information System (INIS)

Bhoje, S.B.; Chellapandi, P.

1996-01-01

Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

Moving ring reactor 'Karin-1'

International Nuclear Information System (INIS)

1983-12-01

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

Preliminary feasibility study of modular reactors

International Nuclear Information System (INIS)

Yamaji, Kenji

1987-01-01

In the future, electric utilities will be required to make a switch-over to a more flexible and dynamic form of power supply due to the slowing growth of power demand, increasing uncertainty, the stagnating economy of increasing scale, the bottleneck of transmission and so on. Nuclear technology would also be required to adapt to this changing environment surrounding its development. The long term prospect of energy demand and nuclear power growth, and the evolution of commercial reactors in Japan are shown. The design of 1,300 MWe advanced LWRs has been completed, and as the reactors of next generation, the ultralarge LWRs of 1,500 - 1,800 MWe are suggested. However, there can be an alternative future for nuclear power development, and in this paper, the possibility for altering the image of conventional nuclear power technology by developing modular reactors which are economical even at small capacity, and can be sited in urban areas just like conventional thermal power plants is examined. The factors for the economical evaluation of modular reactors, learning effect and scale effect on the economy, the case study on a modular high temperature reactor designed by Interatom-GHT, and the possibility of siting in urban areas due to the system of inherent safety are reported. (Kako, I.)

A compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

Neutronic and mechanical design of the reactor core of the Opus system

Energy Technology Data Exchange (ETDEWEB)

Raepsaet, X.; Pascal, S. [CEA Saclay, Dept. Modelisation de Systemes et Structures (DEN/DM2S), 91 - Gif sur Yvette (France)

2007-07-01

Since a few years now, Cea decided to maintain a waking state in its space nuclear activities by carrying out some conceptual studies of embarked nuclear power systems in the range of 100-500 kWe. Results stemming from these ongoing studies are gathered in the project OPUS -Optimized Propulsion Unit System-. This nuclear power system relies on a fast gas-cooled reactor concept coupled either to a Brayton cycle or to a more ambitious energy conversion system using a Hirn cycle to dramatically reduce the size of the radiator. The OPUS reactor core consists of an arrangement of enriched graphite elements of hexagonal cross-section. Their length is equal to the core diameter (48 cm). Coated fuel particles containing enriched (93%) uranium are embedded in these fuel elements. Each fuel element is designed with a centered axial channel through which flows the working fluid: a mixture of helium and xenon gas. This reactor is expected to have an operating life of over 2000 days at full power. In fact the main questions remain on the fuel element manufacturing and on the mechanical design (type and size of particles, packing fraction in the matrix, final core diameter and mass). Especially, the nuclear reactor has been defined considering the possible synergies with the next generation of terrestrial nuclear reactor (International Generation IV Forum). Based on relatively short-term technologies, the same reactor is designed to cover a wide range of power: 100 to 500 kWe without core design modification. The final reactor design presented in this paper is the result of a coupled analysis between the thermomechanical and the neutronic aspects.

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

Data.gov (United States)

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

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

Data.gov (United States)

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

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

International Nuclear Information System (INIS)

Hien, P.D.

1999-01-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

Energy Technology Data Exchange (ETDEWEB)

Hien, P.D. [Vietnam Atomic Energy Agency, Hanoi (Viet Nam)

1999-08-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

Energy Technology Data Exchange (ETDEWEB)

Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

2015-05-15

A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

Utilization of the Dalat Research Reactor for Radioisotope Production, Neutron Activation Analysis, Research and Training

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Duong Van Dong; Cao Dong Vu; Nguyen Xuan Hai

2013-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool type reactor loaded with a mixed core of HEU (36% enrichment) and LEU (19.75% enrichment) fuel assemblies. The reactor is used as a neutron source for the purposes of radioisotopes production, neutron activation analysis, basic and applied research and training. The reactor is operated mainly in continuous runs of 108 hours for cycles of 3–4 weeks for the above mentioned purposes. The current status of safety, operation and utilization of the reactor is given and some aspects for improvement of commercial products and services of the DNRR are also discussed in this paper. (author)

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

Directory of Open Access Journals (Sweden)

Juliana V.F. Sales

2013-01-01

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

Annual progress report for 1982 of Theoretical Reactor Physics Section

International Nuclear Information System (INIS)

Rastogi, B.P.; Kumar, Vinod

1983-01-01

The progress of work done in the Theoretical Reactor Physics Section of the Bhabha Atomic Research Centre, Bombay, during the calendar year 1982 is reported in the form of write-ups and summaries. The main thrust of the work has been to master the neutronic design technology of four different types of nuclear reactor types, namely, pressurized heavy water reactors, boiling light water reactors, pressurized light water reactors and fast breeder reactors. The development work for the neutronic analysis, fuel design, and fuel management of the BWR type reactors of the Tarapur Atomic Power Station has been completed. A new reactor simulator system for PHWR design analysis and core follow-up was completed. Three dimensional static analysis codes based on nodal and finite element methods for the design work of larger size (500-750 MWe) reactors have been developed. Space link kinetics codes in one, two and three dimensions for above-mentioned reactor systems have been written and validated. Fast reactor core disruptive analysis codes have been developed. In the course of R and D work concerning various types of reactor projects, investigations were also carried in the allied areas of Monte Carlo techniques, integral transform methods, path integral methods, high spin states in heavy nuclei and a hydrodynamics model for a laser driven fusion system. (M.G.B.)

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

International Nuclear Information System (INIS)

Rosenthal, Murray Wilford

2009-01-01

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Stress and fatigue analysis of fuelling machine housing of 500 MWe PHWR

International Nuclear Information System (INIS)

Dutta, B.K.; Ramana, W.V.; Kushwaha, H.S.; Kakodkar, A.

1987-01-01

One of the most appealing features of the Pressurised Heavy Water Reactors is the online refuelling capability. For this a fuelling machine is used. This machine opens a reactor channel by removing a seal plug and a shield plug and then does the necessary fuelling by pushing fuel bundles from a fuel magazine by rams. After necessary fuelling the machine closes the channel automatically. One of the most important parts of the fuelling machine is its pressure housing which becomes a part of the reactor channel during refuelling operation. It houses the fuel magazine, separators and rams. Beside channel pressure and other mechanical loads, the pressure housing experiences thermal transients during refuelling. The housing consists of two cylindrical shells having one end-closer in each. They are connected with each other by a large sized coupling. There are many holes on both the end-closers to accommodate ram movement, separators and magazine rive mechanisms. Some of these holes intersect with each other in the housing end-closers and hence end-closers are reinforced accordingly. This also makes the end-closers nonsymmetric. In the following sections the various analysis done to compute general stress distribution, stress concentration factors near to various holes, temperature transients during refuelling and also allowable fatigue cycles for pressure housing of fuelling machine for the proposed 500 MWe are described. (orig.)

Stress and fatigue analysis of fuelling machine housing of 500 MWe PHWR

International Nuclear Information System (INIS)

Dutta, B.K.; Ramana, W.V.; Kushwaha, H.S.; Kakodkar, A.

1987-01-01

One of the most appealing features of the Pressurised Heavy Water Reactors is the online refuelling capability. For this a fuelling machine is used. This machine opens a reactor channel by removing a seal plug and a shield plug and then does the necessary fuelling by pushing fuel bundles from a fuel magazine by rams. After necessary fuelling the machine closes the channel automatically. One of the most important parts of the fuelling machine is its pressure housing which becomes a part of the reactor channel during refuelling operation. It houses the fuel magazine, separators and rams. Beside channel pressure and other mechanical loads, the pressure housing experiences thermal transients during refuelling. The housing consists of two cylindrical shells having one end-closer in each. They are connected with each other by a large sized coupling. There are many holes on both the end-closers to accommodate ram movement, separators and magazine drive mechanisms. Some of these holes intersect with each other in the housing end-closures and hence end-closures are reinforced accordingly. This also makes the end-closures nonsymmetric. In the following sections the various analysis done to compute general stress distribution, stress concentration factors near to various holes, temperature transients during refuelling and also allowable fatigue cycles for pressure housing of fuelling machine for the proposed 500 MWe are described

Compatibility studies of copper, brass and cupronickel with Hytherm-500 thermofluid

International Nuclear Information System (INIS)

Pujar, M.G.; Dayal, R.K.; Gnanamoorthy, J.B.

1989-01-01

Carbon steel used as a structural material in thermofluid/water heat exchangers in the Fast Breeder Test Reactor (FBTR) got perforated resulting in leakage. To suggest an alternative better corrosion resistant material for these exchangers, corrosion resistance studies of copper based alloy systems in both a s received and pickled conditions were carried out in the thermofluid medium (Hytherm-500) at room temperature (∼298 K), 373 K and 423 K upto 500 h duration. The tested materials, copper, admiralty brass and 70/30 cupronickel were found to have excellent corrosion resistance in both as-received and pickled conditions. In all the cases corrosion rates decreased with increased duration of exposure. All the above materials showed better corrosion resistance in pickled condition compared to that in as-received condition. The relative corrosion resistance of these three alloys was as follows: admiralty brass > cop per > cupronickel. This trend in the corrosion resistance was observed in both as-received and pickled conditions. In general, the corrosion resistance in pickled condition was found to be better than that in as-received condition. (author). 3 refs., 3 figs., 2 tabs

28 CFR 54.500 - Employment.

Science.gov (United States)

2010-07-01

... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Employment. 54.500 Section 54.500... in Employment in Education Programs or Activities Prohibited § 54.500 Employment. (a) General. (1) No... subjected to discrimination in employment, or recruitment, consideration, or selection therefor, whether...

Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

International Nuclear Information System (INIS)

Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

1993-01-01

Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

23 CFR 500.109 - CMS.

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2010-04-01

... 23 Highways 1 2010-04-01 2010-04-01 false CMS. 500.109 Section 500.109 Highways FEDERAL HIGHWAY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION TRANSPORTATION INFRASTRUCTURE MANAGEMENT MANAGEMENT AND MONITORING SYSTEMS Management Systems § 500.109 CMS. (a) For purposes of this part, congestion means the level at...

29 CFR 500.0 - Introduction.

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2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Introduction. 500.0 Section 500.0 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR REGULATIONS MIGRANT AND SEASONAL AGRICULTURAL WORKER PROTECTION General Provisions § 500.0 Introduction. (a) The Migrant and Seasonal...

43 CFR 41.500 - Employment.

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2010-10-01

... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Employment. 41.500 Section 41.500 Public... in Employment in Education Programs or Activities Prohibited § 41.500 Employment. (a) General. (1) No... subjected to discrimination in employment, or recruitment, consideration, or selection therefor, whether...

6 CFR 17.500 - Employment.

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2010-01-01

... 6 Domestic Security 1 2010-01-01 2010-01-01 false Employment. 17.500 Section 17.500 Domestic... in Employment in Education Programs or Activities Prohibited § 17.500 Employment. (a) General. (1) No... subjected to discrimination in employment, or recruitment, consideration, or selection therefore, whether...

40 CFR 5.500 - Employment.

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2010-07-01

... 40 Protection of Environment 1 2010-07-01 2010-07-01 false Employment. 5.500 Section 5.500... in Employment in Education Programs or Activities Prohibited § 5.500 Employment. (a) General. (1) No... subjected to discrimination in employment, or recruitment, consideration, or selection therefor, whether...

14 CFR 1253.500 - Employment.

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2010-01-01

... 14 Aeronautics and Space 5 2010-01-01 2010-01-01 false Employment. 1253.500 Section 1253.500... in Employment in Education Programs or Activities Prohibited § 1253.500 Employment. (a) General. (1..., or be subjected to discrimination in employment, or recruitment, consideration, or selection therefor...

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-15

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

A review of fast reactor programme in India - April 1992

Energy Technology Data Exchange (ETDEWEB)

Paranjpe, S R [Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

1992-07-01

There is no change in the basic policy for development of nuclear energy in India. Fast Breeder Reactors are required to be available commercially to supply increasing quantities of nuclear energy when the first phase programme of deployment of Pressurised Heavy Water Reactors would be reaching the limit imposed by indigenously available natural uranium. Based on presently proven reserves of economically exploitable uranium one cannot expect to support more than 10 to 15 million kilowatt of installed capacity of PHWRs. The immediate goal of the Fast Reactor Programme therefore, remains completion by 2002-2003 of the first 500 MWe Prototype Fast Breeder Reactor which will become the first reactor in the series of reactors to be built there afterwards. This will enable addition of one 500 MWe reactor each year even if the first phase of programme of PHWR is limited to 6.0 million kilowatt. The capital cost of installed kilowatt for FBRs is expected to be comparable to the capital cost per kilowatt for PHWRS. It is expected to launch the construction of PFBR in the next 2 or 3 years as soon as the over all economic condition shows some improvement. In the meantime, manufacturing development of important NSS components like Steam Generators, Sodium Pumps, Main Vessel and Inner Vessel has been initiated. Detailed designs of Control Rod Drive Mechanism (Primary) has been completed and contacts with the manufacturers are being established to identify the industry which would be entrusted with the responsibility of manufacturing the Control Rod Drive Mechanisms. Manufacturing technology for making cladding tubes of D9 stainless steel has been developed and significant progress has been made towards the production of hexagonal wrapper (i.e. Hex-Cans). Inclined Fuel Transfer Machine for loading and unloading the fuel from the Main Vessel has been designed and manufacturing of the prototype machine has been initiated. It is hoped that these steps will enable timely completion

Physical Characteristics of the Dalat Nuclear Research Reactor; Cac dac trung vat ly lo cua lo phan ung hat nhan Da Lat

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [ed.; Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor.

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

Directory of Open Access Journals (Sweden)

Jeél Moya S

2015-12-01

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

21 CFR 556.500 - Oxytetracycline.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Oxytetracycline. 556.500 Section 556.500 Food and... Residues of New Animal Drugs § 556.500 Oxytetracycline. (a) Acceptable daily intake (ADI). The ADI for total tetracycline residues (chlortetracycline, oxytetracycline, and tetracycline) is 25 micrograms per...

29 CFR 500.155 - Authority.

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2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Authority. 500.155 Section 500.155 Labor Regulations... AGRICULTURAL WORKER PROTECTION Enforcement Agreements with Federal and State Agencies § 500.155 Authority...) to Federal and State agencies such authority (other than rulemaking) as he determines may be useful...

1 CFR 500.140 - Employment.

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2010-01-01

... 1 General Provisions 1 2010-01-01 2010-01-01 false Employment. 500.140 Section 500.140 General... ACTIVITIES CONDUCTED BY THE NATIONAL COMMISSION FOR EMPLOYMENT POLICY § 500.140 Employment. No qualified handicapped person shall, on the basis of handicap, be subjected to discrimination in employment under any...

22 CFR 229.500 - Employment.

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2010-04-01

... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Employment. 229.500 Section 229.500 Foreign... OR ACTIVITIES RECEIVING FEDERAL FINANCIAL ASSISTANCE Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 229.500 Employment. (a) General. (1) No person shall...

22 CFR 146.500 - Employment.

Science.gov (United States)

2010-04-01

... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Employment. 146.500 Section 146.500 Foreign... ACTIVITIES RECEIVING FEDERAL FINANCIAL ASSISTANCE Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 146.500 Employment. (a) General. (1) No person shall, on the...

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)

Fatigue cycles evaluation of 500 MWe PHWR coolant channel sealdisc

International Nuclear Information System (INIS)

Chawla, D.S.; Vaze, K.K.; Kushwaha, H.S.; Gupta, K.S.; Bhambra, H.S.

1998-07-01

At each end of coolant channel there is one sealing plug assembly. The sealdisc is a part of sealing plug assembly. The sealdisc is used to avoid leakage of heavy water. The importance of sealdisc can be understood by the fact that there are 784 sealdiscs in one 500 MWe PHWR unit. During the life time of reactor the sealdisc will be subjected to cyclic loads due to reactor startup, shutdown, power setback and also due to refuelling operations. Excessive reversal of stresses may lead to fatigue failure. The sealdisc failure may cause loss of coolant accidents. Since sealdisc is safety class 1 component, it has to be qualified according to ASME Section III Division 1 NB. For cyclic loads, the fatigue analysis is essential to assess the allowable number of cycles and also to check the total usage factor due to different cyclic loads. To evaluate the allowable fatigue cycles, the analysis is carried out using finite element method. The present report deals with the fatigue cycles evaluation of 500 MWe PHWR sealdisc. The finite element model having eight noded axisymmetric elements is used for the analysis. The various loads considered in the analysis are mechanical loads arising due to refuelling operations and number of temperature-pressure transients. During refuelling, the sealdisc is removed and reinstalled back by use of fuelling machine ram which applies load at centre as well as at rocker point of sealdisc. The stress analysis is carried out for each stage of loading during refuelling and fatigue cycles are evaluated. For temperature transient, decoupled thermal analysis is carried out. At various instants of time, the stresses are computed using temperatures calculated in thermal analysis. The pressure variation is also considered along with temperature variation. The fatigue cycles are evaluated for each transient using maximum alternating stress intensities. The usage factors are calculated for various temperature/pressure transients and refuelling loads

16 CFR 500.17 - Fractions.

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2010-01-01

... 16 Commercial Practices 1 2010-01-01 2010-01-01 false Fractions. 500.17 Section 500.17 Commercial... LABELING ACT § 500.17 Fractions. (a) SI metric declarations of net quantity of contents of any consumer commodity may contain only decimal fractions. Other declarations of net quantity of contents may contain...

7 CFR 500.6 - Gambling.

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2010-01-01

... 7 Agriculture 6 2010-01-01 2010-01-01 false Gambling. 500.6 Section 500.6 Agriculture Regulations... NATIONAL ARBORETUM Conduct on U.S. National Arboreturm Property § 500.6 Gambling. Participating in games for money or other personal property, or the operation of gambling devices, the conduct of a lottery...

9 CFR 93.500 - Definitions.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Definitions. 93.500 Section 93.500... CONVEYANCE AND SHIPPING CONTAINERS Swine § 93.500 Definitions. Wherever in this subpart the following terms..., Spain, Sweden, and the United Kingdom (England, Scotland, Wales, the Isle of Man, and Northern Ireland...

21 CFR 500.82 - Definitions.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Definitions. 500.82 Section 500.82 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS, AND RELATED PRODUCTS GENERAL Regulation of Carcinogenic Compounds Used in Food-Producing Animals § 500...

29 CFR 500.162 - Reports.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Reports. 500.162 Section 500.162 Labor Regulations Relating... AGRICULTURAL WORKER PROTECTION Enforcement Agreements with Federal and State Agencies § 500.162 Reports. The Secretary shall require such reports as he deems necessary of activities conducted pursuant to State plans...

24 CFR 180.500 - Discovery.

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2010-04-01

... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Discovery. 180.500 Section 180.500... OPPORTUNITY CONSOLIDATED HUD HEARING PROCEDURES FOR CIVIL RIGHTS MATTERS Discovery § 180.500 Discovery. (a) In general. This subpart governs discovery in aid of administrative proceedings under this part. Discovery in...

13 CFR 500.105 - Staff.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Staff. 500.105 Section 500.105... LOAN PROGRAM Board Procedures § 500.105 Staff. (a) Executive Director. The Executive Director of the... direction with respect to the administration of the Board's actions, directs the activities of the staff...

10 CFR 1042.500 - Employment.

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2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Employment. 1042.500 Section 1042.500 Energy DEPARTMENT OF... RECEIVING FEDERAL FINANCIAL ASSISTANCE Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 1042.500 Employment. (a) General. (1) No person shall, on the basis of...

45 CFR 618.500 - Employment.

Science.gov (United States)

2010-10-01

... 45 Public Welfare 3 2010-10-01 2010-10-01 false Employment. 618.500 Section 618.500 Public Welfare... the Basis of Sex in Employment in Education Programs or Activities Prohibited § 618.500 Employment. (a... benefits of, or be subjected to discrimination in employment, or recruitment, consideration, or selection...

Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap

International Nuclear Information System (INIS)

Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

1994-01-01

The nominal power of the reconstructed Dalat reactor is of 500 KW. After a meticulous preparation the Russian and Vietnamese teams have proceeded to the physical reactor start-up in November 1983 with the core configuration exempt from the neutron trap. The reactor has reached the physical criticality at 19h50 on 1 November 1983. The report delineates different steps of the start-up procedure. 2 refs., 3 figs., 7 tabs

Review of fast reactor activities in India

Energy Technology Data Exchange (ETDEWEB)

Paranjpe, S R [Reactor Research Centre, Kalpakkam, Tamil Nadu (India)

1981-05-01

It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties.

Review of fast reactor activities in India

International Nuclear Information System (INIS)

Paranjpe, S.R.

1981-01-01

It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties

24 CFR 214.500 - Audit.

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 2 2010-04-01 2010-04-01 false Audit. 214.500 Section 214.500... PROGRAM Other Federal Requirements § 214.500 Audit. Housing counseling grant recipients and subrecipients shall be subject to the audit requirements contained in 24 CFR parts 84 and 85. HUD must be provided a...

10 CFR 5.500 - Employment.

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2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Employment. 5.500 Section 5.500 Energy NUCLEAR REGULATORY... FINANCIAL ASSISTANCE Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 5.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from...

32 CFR 196.500 - Employment.

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2010-07-01

... 32 National Defense 2 2010-07-01 2010-07-01 false Employment. 196.500 Section 196.500 National... Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 196.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from participation in, be...

45 CFR 2555.500 - Employment.

Science.gov (United States)

2010-10-01

... 45 Public Welfare 4 2010-10-01 2010-10-01 false Employment. 2555.500 Section 2555.500 Public... Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 2555.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from participation in, be...

Engineering Design of a Double Reactor for Spent Fuel Oxidation

International Nuclear Information System (INIS)

Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee

2015-01-01

In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more

Engineering Design of a Double Reactor for Spent Fuel Oxidation

Energy Technology Data Exchange (ETDEWEB)

Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more.

23 CFR 500.104 - State option.

Science.gov (United States)

2010-04-01

... 23 Highways 1 2010-04-01 2010-04-01 false State option. 500.104 Section 500.104 Highways FEDERAL HIGHWAY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION TRANSPORTATION INFRASTRUCTURE MANAGEMENT MANAGEMENT AND MONITORING SYSTEMS Management Systems § 500.104 State option. Except as specified in § 500.105 (a...

Nordic study on reactor waste

International Nuclear Information System (INIS)

1981-08-01

In 1981, 14 nuclear power reactors are in operation and 2 under construction in the Nordic countries. So far, the reactor waste originating from day-to-day operation of these plants has been stored in solidified form at the reactor sites. Within a few years a satisfactory disposal procedure needs to be established. While the main R and D effects in the waste field have earlier been devoted to the question of irradiated fuel and waste from reprocessing, there is therefore now an increased interest in reactor waste with its much lower radioactivity but somewhat larger volumes. Since 1977, efforts have been made in a joint Nordic study to examine which facts need to be known in order to perform a comprehensive safety assessment of a reactor waste management system. In the present study a Reference system related to the waste generated over 30 years from six 500 MW-reactors is examined. The dominating radionuclides during storage and transportation accident scenarios are Cs-134, Cs-137 and Co-60. For most of the release scenarios from repositories Cs-137 and Sr-90 are dominating. Some scenarios are, however, dominated by the very longlived nuclides I-129 and C-14. A closer examination of the concentration in the waste of these nuclides and of their leaching properties indicates that their small - but significant - influence, as calculated, is probably grossly overestimated. The mechanical stability obtained in routine solidification processes of reactor waste products in conjunction with the outer container (steel drum, transport container, etc.) turns out to be sufficient. Difficulties were encountered in applying ICRP methodology and available dose calculation methods to calculation of population doses due to small activity releases, and effects extending into the far future. (EG)

Current status of operation, utilization and refurbishment of the Dalat nuclear research reactor

International Nuclear Information System (INIS)

Pham Duy Hien.

1993-01-01

The reconstructed nuclear research reactor at Dalat, Vietnam has been put into operation since March 1984. Up to present a cumulative operation time of 13,172 hrs at nominal power (500 kW) has been recorded. Production of radioisotopes for medical uses, element analysis by using activation techniques, as well as fundamental and applied research with filtered neutrons are the main activities of reactor utilizations. The problems facing Dalat Nuclear Research Institute are the ageing of the re-used TRIGA-MARK-II reactor components (especially the corrosion of the reactor tank), as well as the obsolescence of many equipment and components of the reactor control and instrumentation system. Refurbishment works are being in process with the technical and financial supports from the Vietnam government and the IAEA. (author). 7 refs, 2 tabs, 10 figs

Fast Pyrolysis of Lignin Using a Pyrolysis Centrifuge Reactor

DEFF Research Database (Denmark)

Trinh, Ngoc Trung; Jensen, Peter Arendt; Sárossy, Zsuzsa

2013-01-01

Fast pyrolysis of lignin from an ethanol plant was investigated on a lab scale pyrolysis centrifuge reactor (PCR) with respect to pyrolysis temperature, reactor gas residence time, and feed rate. A maximal organic oil yield of 34 wt % dry basis (db) (bio-oil yield of 43 wt % db) is obtained...... at temperatures of 500−550 °C, reactor gas residence time of 0.8 s, and feed rate of 5.6 g/min. Gas chromatography mass spectrometry and size-exclusion chromatography were used to characterize the Chemical properties of the lignin oils. Acetic acid, levoglucosan, guaiacol, syringols, and p-vinylguaiacol are found...... components and molecular mass distribution of the lignin oils. The obtained lignin oil has a very different components composition when compared to a beech wood oil....

24 CFR 3.500 - Employment.

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Employment. 3.500 Section 3.500... Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 3.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from participation in, be...

Design of High Temperature Reactor Vessel Using ANSYS Software

International Nuclear Information System (INIS)

Bandriyana; Kasmudin

2003-01-01

Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

Instrumentation and control of future sodium cooled fast reactors - Design improvements

International Nuclear Information System (INIS)

Madhusoodanan, K.; Sakthivel, M.; Chellapandi, P.

2013-06-01

India's fast reactor program started with the 40 MWt Fast Breeder Test Reactor. 500 MWe Prototype Fast Breeder Reactor (PFBR) is currently under construction at Kalpakkam. Safety of PFBR is enhanced by improved design features of I and C system. Since the design of Instrumentation and control (I and C) of PFBR, considerable improvements in terms of advancement in technology and indigenization has taken place. Further improvements in I and C is proposed for solving many of the difficulties faced during the design and construction phases of PFBR. Design improvements proposed are covered in this paper which will make the implementation and maintenance of I and C of future SFRs easier. (authors)

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

1994-01-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

Problems in the assessment of inherent safety characteristics of nuclear reactors

International Nuclear Information System (INIS)

Garribba, S.F.; Vivante, C.

1988-01-01

A number of proposals are being made for an increased RD and D effort on advanced nuclear power reactors that would display outstanding safety performance. A common characteristic of the different reactor concepts would be their limited reliance upon active engineered systems under major accident conditions. However, when submitted to a more close scrutiny reactor concept options may reveal diverging safety behaviors and also development opportunities. In this respect, three issues are explored in this paper. A first question is the meaning of non-active, i.e. inherent and passive safety features. Next, is the ranking of advanced and new reactor concepts from the viewpoint of inherent and passive safety. Multiple correspondence analysis may provide a simple tool, whose use is shown for the case of HTR-500, AP600 and PRISM. Conversely, probabilistic risk assessment would allow quantitative comparisons, although lack of information and data is an obstacle. Finally, is demonstration of safety performances as a step toward market deployment of the new reactor systems

Advanced Nuclear Reactor Concepts for China

International Nuclear Information System (INIS)

Knoche, D.; Sassen, F.; Tietsch, W.; Yujie, Dong; Li, Cao

2008-01-01

China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

Advanced Nuclear Reactor Concepts for China

Energy Technology Data Exchange (ETDEWEB)

Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

2008-07-01

China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

Testing of a reactimeter for a light water reactor in the range + 500 to - 5000 pcm; Essai d'un reactimetre pour reacteur a eau legere dans la gamme + 500, - 5000 pcm

Energy Technology Data Exchange (ETDEWEB)

Chauvet, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

This apparatus is designed to measure instantaneously the positive or negative reactivity of a uranium reactor moderated by light water, on condition that the point of departure is the critical state of the reactor, or an already known sub-critical state. Slight modifications only are required to adapt it to another type of reactor. It is an analogue computer which simply inverses the transfer function of the reactor; it is not therefore a model reactor of which the output voltage is connected by a servo-mechanism to the power of the reactor to give the reactivity; the principle of the calculation of the reactivity does not depend on a servomechanism. One of its disadvantages is that it cannot operate outside a power variation range of 2.5 decades. However the measurement of a negative reactivity value between 0 and 3000 pcm is immediate. It measures the reactivity without deducting it from the period; it therefore gives the reactivity very precisely both for divergence and convergence even through in this latter case the period does not in fact exist. The equipment makes it possible to calibrate very rapidly the control rods of a reactor (the rod-drop method), to measure the reactivity of an experiment in the core, and to measure certain temperature effects. It is also possible by introducing a control into the core at a measured rate, to deduce directly its efficiency curve. (author) [French] Cet appareil est destine a mesurer instantanement la reactivite positive ou negative d'un reacteur a uranium modere a l'eau legere, a condition de partir de l'etat critique du reacteur, ou eventuellement d'un etat sous-critique deja connu. De legeres modifications permettent de l'adapter a un autre type de moderateur. C'est un calculateur analogique, qui inverse purement et simplement la fonction de transfert du reacteur; ce n'est donc pas un simulateur de pile dont la tension de sortie est asservie a la puissance du reacteur pour elaborer la reactivite; le principe du

Determination of the lowest critical power levels of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Binh, Do Quang; Nghiem, Huynh Ton; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

This paper presents the experimental methods for determining critical states of the Dalat Nuclear Research Reactor containing an extraneous neutron source induced by gamma ray reactions on beryllium in the reactor. The lowest critical power levels are measured at various moments after the reactor is shut down following 100 hours of its continuous operation. Th power levels vary from (0.5-1.2) x 10{sup -4} of P{sub n}, i.e. (25-60)W to (1.1-1.6) x 10{sup -5} of P{sub n}, i.e. (5.5-8)W at corresponding times of 4 days to 13 days after the reactor is shut down. However the critical power must be chosen greater than 500 W to sustain the steady criticality of the reactor for a long time. (author). 3 refs. 4 figs. 1 tab.

Value addition initiatives for CANDU reactor operation performance

International Nuclear Information System (INIS)

Chugh, V.; Parmar, R.; Schut, J.; Sherin, J.; Xie, H.; Zobin, D.

2013-01-01

Recently, AMEC NSS initiated projects for CANDU® station performance engineering with potentially high returns for the utilities. This paper discusses three initiatives. Firstly, optimization of instrument calibration interval from 1 to 3 years will reduce time commitments on the maintenance resources on top of financial savings ~$3,500 per instrument. Secondly, reactor thermal power uncertainty assessment shows the level of operation which is believed to have an over-conservative margin that can be used to increase power by up to 0.75%. Finally, as an alternative means for controlling Reactor Inlet Header Temperature (RIHT), physical modifications to the High Pressure (HP) feedwater heaters can be useful for partially recovering RIHT resulting in increased production by 10-12 MWe. (author)

The launching of the construction of the Jules Horowitz Reactor

International Nuclear Information System (INIS)

Anon.

2007-01-01

In March 2007 the French deputy minister of industry has officially launched the construction of the new research reactor called Jules Horowitz (RJH) on the Cea site of Cadarache. RJH, that is due to operate in 2014, will be used to study the aging process of irradiated materials in any type of reactors, the behaviour of new nuclear fuels irradiated in different configurations and scenarios, and to produce radionuclides for nuclear medicine and high-quality doped silicon for the electronics industry. The investment that reaches 500 million euros is dispatched between Cea (50%), EDF (20%), Areva (10%) and foreign contributors (20%). (A.C.)

29 CFR 500.9 - Discrimination prohibited.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Discrimination prohibited. 500.9 Section 500.9 Labor... SEASONAL AGRICULTURAL WORKER PROTECTION General Provisions § 500.9 Discrimination prohibited. (a) It is a... Secretary alleging such discrimination. ...

Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

Directory of Open Access Journals (Sweden)

Kaiser Krista

2016-01-01

Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

Flow analysis in a supercritical water oxidation reactor

International Nuclear Information System (INIS)

Oh, C.H.; Kochan, R.J.; Beller, J.M.

1996-01-01

Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Rosenthal, Murray Wilford [ORNL

2009-08-01

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Dien, Nguyen Nhi; Lam, Pham Van; Phuong, Huynh Dong; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10{sup 5}/10{sup 8} n/cm{sup 2}/sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to ({gamma},n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is {beta}{sup B}e{sub eff}=0.49%{beta}{sub eff} for a beryllium weight relative to U{sup 235} fuel of m{sub B}e/m{sub U} = 8.5. This result is acceptable in comparison to those obtained for other Be-U{sup 235} media. (author). 5 refs., 2 figs., 4 tabs.

Liquid metal coolant disposal from UKAEA reactors at Dounreay

International Nuclear Information System (INIS)

Adam, E.R.

1997-01-01

As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

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

Directory of Open Access Journals (Sweden)

Michael V. Zaragoza

2017-11-01

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

Solid State Reactor Final Report

Energy Technology Data Exchange (ETDEWEB)

Mays, G.T.

2004-03-10

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

Time history of diesel particle deposition in cylindrical dielectric barrier discharge reactors

Energy Technology Data Exchange (ETDEWEB)

Talebizadeh, P.; Rahimzadeh, H., E-mail: rahimzad@aut.ac.ir [Amirkabir University of Technology, Department of Mechanical Engineering (Iran, Islamic Republic of); Ahmadi, G. [Clarkson University, Department of Mechanical and Aeronautical Engineering (United States); Brown, R. [Queensland University of Technology, Biofuel Engine Research Facility (Australia); Inthavong, K. [RMIT University, School of Aerospace, Mechanical and Manufacturing Engineering (Australia)

2016-12-15

Non-thermal plasma (NTP) treatment reactors have recently been developed for elimination of diesel particulate matter for reducing both the mass and number concentration of particles. The role of the plasma itself is obscured by the phenomenon of particle deposition on the reactor surface. Therefore, in this study, the Lagrangian particle transport model is used to simulate the dispersion and deposition of nano-particles in the range of 5 to 500 nm in a NTP reactor in the absence of an electric field. A conventional cylindrical dielectric barrier discharge reactor is selected for the analysis. Brownian diffusion, gravity and Saffman lift forces were included in the simulations, and the deposition efficiencies of different sized diesel particles were studied. The results show that for the studied particle diameters, the effect of Saffman lift is negligible and gravity only affects the motion of particles with a diameter of 500 nm or larger. Time histories of particle transport and deposition were evaluated for one-time injection and a continuous (multiple-time) injection. The results show that the number of deposited particles for one-time injection is identical to the number of deposited particles for multiple-time injections when adjusted with the shift in time. Furthermore, the maximum number of escaped particles occurs at 0.045 s after the injection for all particle diameters. The presented results show that some particle reduction previously ascribed to plasma treatment has ignored contributions from the surface deposition.

Time history of diesel particle deposition in cylindrical dielectric barrier discharge reactors

International Nuclear Information System (INIS)

Talebizadeh, P.; Rahimzadeh, H.; Ahmadi, G.; Brown, R.; Inthavong, K.

2016-01-01

Non-thermal plasma (NTP) treatment reactors have recently been developed for elimination of diesel particulate matter for reducing both the mass and number concentration of particles. The role of the plasma itself is obscured by the phenomenon of particle deposition on the reactor surface. Therefore, in this study, the Lagrangian particle transport model is used to simulate the dispersion and deposition of nano-particles in the range of 5 to 500 nm in a NTP reactor in the absence of an electric field. A conventional cylindrical dielectric barrier discharge reactor is selected for the analysis. Brownian diffusion, gravity and Saffman lift forces were included in the simulations, and the deposition efficiencies of different sized diesel particles were studied. The results show that for the studied particle diameters, the effect of Saffman lift is negligible and gravity only affects the motion of particles with a diameter of 500 nm or larger. Time histories of particle transport and deposition were evaluated for one-time injection and a continuous (multiple-time) injection. The results show that the number of deposited particles for one-time injection is identical to the number of deposited particles for multiple-time injections when adjusted with the shift in time. Furthermore, the maximum number of escaped particles occurs at 0.045 s after the injection for all particle diameters. The presented results show that some particle reduction previously ascribed to plasma treatment has ignored contributions from the surface deposition.

Considerations in providing purification flows for 500 MWe PHWR primary circuits

International Nuclear Information System (INIS)

Sharma, A.K.; Goswami, S.; Bapat, C.N.; Sharma, V.K.

1995-01-01

The purpose of the purification system is to keep the primary heat transport (PHT) system clean by removing traces of impurities arising due to corrosion of the carbon steel pipes and heat transfer surfaces and erosion/corrosion of valve trims, pipes and mechanical seals or due to presence of soluble or insoluble fission products. These impurities are undesirable because they are usually radioactive, either naturally or through activation by the neutron flux as they are carried by the coolant through the reactor core. The purification system minimizes the probability of generation of radioactive impurities by controlling the chemistry of PHT coolant so that corrosion is minimum. Various considerations for providing the requisite purification flow to fulfill the above functions for a typical 500 MWe PHWR are presented. (author). 4 refs., 2 tabs., 2 figs

Secondary cycle water chemistry for 500 MWe pressurised heavy water reactor (PHWR) plant: a case study

International Nuclear Information System (INIS)

Bhandakkar, A.; Subbarao, A.; Agarwal, N.K.

1995-01-01

In turbine and secondary cycle system of 500 MWe PHWR, chemistry of steam and water is controlled in secondary cycle for prevention of corrosion in steam generators (SGs), feedwater system and steam system, scale and deposit formation on heat transfer surfaces and carry-over of solids by steam and deposition on steam turbine blades. Water chemistry of secondary side of SGs and turbine cycle is discussed. (author). 8 refs., 2 tabs., 1 fig

TREATMENT OF REFRACTORY OXIDES IN HF-PLASMA REACTORS

OpenAIRE

Bakhvalov , A.; Dresvin , S.; Levitskaya , T.; Paskalov , G.; Philippov , A.

1990-01-01

Results of theoretical and experimental studies of SiO2 NaBSi, MgO, W and some other materials treatment in induction type high-frequency plasma under atmospheric pressure are presented. Key study objective - optimization of plasma installation operating modes with maximum efficiency -0.6 -0.7 ; spheroidization extent -90-99%, size of treated particles 1-500 mkm. Diagnostics of thermophysical and gasodynamical plasma reactor specifications has been presented.

1 CFR 500.110 - Self-evaluation.

Science.gov (United States)

2010-01-01

... 1 General Provisions 1 2010-01-01 2010-01-01 false Self-evaluation. 500.110 Section 500.110... PROGRAMS OR ACTIVITIES CONDUCTED BY THE NATIONAL COMMISSION FOR EMPLOYMENT POLICY § 500.110 Self-evaluation... or organizations representing handicapped persons, to participate in the self-evaluation process by...

The optimal control of ITU TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Can, Burhanettin

2008-01-01

In this study, optimal control of ITU TRIGA Mark-II Reactor is discussed. A new controller has been designed for ITU TRIGA Mark-II Reactor. The controller consists of main and auxiliary controllers. The form is based on Pontragyn's Maximum Principle and the latter is based on PID approach. For the desired power program, a cubic function is chosen. Integral Performance Index includes the mean square of error function and the effect of selected period on the power variation. YAVCAN2 Neutronic - Thermal -Hydraulic code is used to solve the equations, namely 11 equations, dealing with neutronic - thermal - hydraulic behavior of the reactor. For the controller design, a new code, KONTCAN, is written. In the application of the code, it is seen that the controller controls the reactor power to follow the desired power program. The overshoot value alters between 100 W and 500 W depending on the selected period. There is no undershoot. The controller rapidly increases reactivity, then decreases, after that increases it until the effect of temperature feedback is compensated. Error function varies between 0-1 kW. (author)

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

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

Directory of Open Access Journals (Sweden)

Sônia Terezinha dos Anjos Lopes

1993-12-01

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

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

CERN Document Server

Jovicevic, Jelena; Strandberg, Jonas

2015-11-20

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

47 CFR 65.500 - Net income.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 3 2010-10-01 2010-10-01 false Net income. 65.500 Section 65.500... OF RETURN PRESCRIPTION PROCEDURES AND METHODOLOGIES Interexchange Carriers § 65.500 Net income. The net income methodology specified in § 65.450 shall be utilized by all interexchange carriers that are...

20 CFR 638.500 - Orientation program.

Science.gov (United States)

2010-04-01

... 20 Employees' Benefits 3 2010-04-01 2010-04-01 false Orientation program. 638.500 Section 638.500 Employees' Benefits EMPLOYMENT AND TRAINING ADMINISTRATION, DEPARTMENT OF LABOR JOB CORPS PROGRAM UNDER TITLE IV-B OF THE JOB TRAINING PARTNERSHIP ACT Center Operations § 638.500 Orientation program. The...

Neutron radiation damage studies in the structural materials of a 500 MWe fast breeder reactor using DPA cross-sections from ENDF / B-VII.1

Science.gov (United States)

Saha, Uttiyoarnab; Devan, K.; Bachchan, Abhitab; Pandikumar, G.; Ganesan, S.

2018-04-01

The radiation damage in the structural materials of a 500 MWe Indian prototype fast breeder reactor (PFBR) is re-assessed by computing the neutron displacement per atom (dpa) cross-sections from the recent nuclear data library evaluated by the USA, ENDF / B-VII.1, wherein revisions were taken place in the new evaluations of basic nuclear data because of using the state-of-the-art neutron cross-section experiments, nuclear model-based predictions and modern data evaluation techniques. An indigenous computer code, computation of radiation damage (CRaD), is developed at our centre to compute primary-knock-on atom (PKA) spectra and displacement cross-sections of materials both in point-wise and any chosen group structure from the evaluated nuclear data libraries. The new radiation damage model, athermal recombination-corrected displacement per atom (arc-dpa), developed based on molecular dynamics simulations is also incorporated in our study. This work is the result of our earlier initiatives to overcome some of the limitations experienced while using codes like RECOIL, SPECTER and NJOY 2016, to estimate radiation damage. Agreement of CRaD results with other codes and ASTM standard for Fe dpa cross-section is found good. The present estimate of total dpa in D-9 steel of PFBR necessitates renormalisation of experimental correlations of dpa and radiation damage to ensure consistency of damage prediction with ENDF / B-VII.1 library.

Membrane steam reforming of natural gas for hydrogen production by utilization of medium temperature nuclear reactor

International Nuclear Information System (INIS)

Djati Hoesen Salimy

2010-01-01

The assessment of steam reforming process with membrane reactor for hydrogen production by utilizing of medium temperature nuclear reactor has been carried out. Difference with the conventional process of natural gas steam reforming that operates at high temperature (800-1000°C), the process with membrane reactor operates at lower temperature (~500°C). This condition is possible because the use of perm-selective membrane that separate product simultaneously in reactor, drive the optimum conversion at the lower temperature. Besides that, membrane reactor also acts the role of separation unit, so the plant will be more compact. From the point of nuclear heat utilization, the low temperature of process opens the chance of medium temperature nuclear reactor utilization as heat source. Couple the medium temperature nuclear reactor with the process give the advantage from the point of saving fossil fuel that give direct implication of decreasing green house gas emission. (author)

Unification of reactor elastomeric sealing based on material

International Nuclear Information System (INIS)

Sinha, N.K.; Raj, Baldev

2012-01-01

The unification of elastomeric sealing applications of Indian nuclear reactors based on a few qualified fluoroelastomer/perfluoroelastomer compounds and standardized approaches for finite element analysis (FEA) based design, manufacturing process and antifriction coatings is discussed. It is shown that the advance polymer architecture based Viton ® formulation developed for inflatable seals of 500 MWe Prototype Fast Breeder Reactor (PFBR) and its four basic variations can encompass other sealing applications of PFBR with minimum additional efforts on development and validation. Changing the blend ratio of Viton ® GBL 200S and 600S in inflatable seal formulation could extend its use to Pressurized Heavy Water Reactors (PHWRs). The higher operating temperature of Advanced Heavy Water Reactor (AHWR) seals expands the choice to perfluoroelastomers. FEA based on plane-strain/axisymmetric modeling (with Mooney–Rivlin as the basic constitutive model), seal manufacture by cold feed extrusion and injection molding as well as plasma Teflon-like coating belonging to two variations obtained from the development of inflatable seals provide the necessary standardization for unification. The gains in simplification of design, development and operation of seals along with the enhancements of safety and reliability are expected to be substantial.

RMB. The new Brazilian multipurpose research reactor

International Nuclear Information System (INIS)

Perrotta, Jose Augusto; Soares, Adalberto Jose

2015-01-01

Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also presents the

RMB. The new Brazilian multipurpose research reactor

Energy Technology Data Exchange (ETDEWEB)

Perrotta, Jose Augusto; Soares, Adalberto Jose [Comissao Nacional de Energia Nuclear (CNEN) (Brazil)

2015-01-15

Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also

13 CFR 500.202 - Loan amount.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Loan amount. 500.202 Section 500.202 Business Credit and Assistance EMERGENCY OIL AND GAS GUARANTEED LOAN BOARD EMERGENCY OIL AND GAS GUARANTEED LOAN PROGRAM Oil and Gas Guaranteed Loans § 500.202 Loan amount. The aggregate amount of loan...

31 CFR 500.301 - Foreign country.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Foreign country. 500.301 Section 500.301 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... § 500.301 Foreign country. The term foreign country also includes, but not by way of limitation: (a) The...

31 CFR 500.412 - Process vs. manufacture.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Process vs. manufacture. 500.412 Section 500.412 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE... Interpretations § 500.412 Process vs. manufacture. A commodity subject to § 500.204 remains subject howsoever it...

Nuclear engineering laboratory self regulated power oscillation experiments at the Health Physics Research Reactor

International Nuclear Information System (INIS)

Miller, L.F.; Mihalczo, J.T.; Bailiff, E.G.; Woody, N.D.; Gardner, G.D.

1983-01-01

Self regulated power oscillation experiments with a variety of initial conditions have been performed with the ORNL Health Physics Research Reactor (HPRR) by undergraduate nuclear engineering students from The University of Tennessee for several years. These experiments demonstrate the coupling between reactor kinetics and heat transfer and show how the temperature coefficient of reactivity affects reactor behavior. A model that consists of several coupled first order nonlinear differential equations is used to calculate the temperature of the core center and surface and power as a function of time which are compared with the experimental data; also, the model is also used to study the effects of various model parameters and initial conditions on the amplitude, frequency and damping of the power and temperature oscillations. A previous paper presented some limited experimental results and demonstrated the correspondence between a simple point model and the experimental data. This paper presents the results of experiments for: (1) the initial power fixed at 9 kW with central core temperatures of 300 0 F and 500 0 F, annd (2) the initial central core temperature fixed at 500 0 F with initial powers of 6 and 8 kW

Nuclear vapor thermal reactor propulsion technology

International Nuclear Information System (INIS)

Maya, I.; Diaz, N.J.; Dugan, E.T.; Watanabe, Y.; McClanahan, J.A.; Wen-Hsiung Tu; Carman, R.L.

1993-01-01

The conceptual design of a nuclear rocket based on the vapor core reactor is presented. The Nuclear Vapor Thermal Rocket (NVTR) offers the potential for a specific impulse of 1000 to 1200 s at thrust-to-weight ratios of 1 to 2. The design is based on NERVA geometry and systems with the solid fuel replaced by uranium tetrafluoride (UF 4 ) vapor. The closed-loop core does not rely on hydrodynamic confinement of the fuel. The hydrogen propellant is separated from the UF 4 fuel gas by graphite structure. The hydrogen is maintained at high pressure (∼100 atm), and exits the core at 3,100 K to 3,500 K. Zirconium carbide and hafnium carbide coatings are used to protect the hot graphite from the hydrogen. The core is surrounded by beryllium oxide reflector. The nuclear reactor core has been integrated into a 75 klb engine design using an expander cycle and dual turbopumps. The NVTR offers the potential for an incremental technology development pathway to high performance gas core reactors. Since the fuel is readily available, it also offers advantages in the initial cost of development, as it will not require major expenditures for fuel development

13 CFR 500.204 - Loan terms.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Loan terms. 500.204 Section 500.204 Business Credit and Assistance EMERGENCY OIL AND GAS GUARANTEED LOAN BOARD EMERGENCY OIL AND GAS GUARANTEED LOAN PROGRAM Oil and Gas Guaranteed Loans § 500.204 Loan terms. (a) All loans guaranteed under the...

Technology development of fast reactor fuel reprocessing technology in India

International Nuclear Information System (INIS)

Natarajan, R.; Raj, Baldev

2009-01-01

India is committed to the large scale induction of fast breeder reactors beginning with the construction of 500 MWe Prototype Fast Breeder Reactor, PFBR. Closed fuel cycle is a prerequisite for the success of the fast reactors to reduce the external dependence of the fuel. In the Indian context, spent fuel reprocessing, with as low as possible out of pile fissile inventory, is another important requirement for increasing the share in power generation through nuclear route as early as possible. The development of this complex technology is being carried out in four phases, the first phase being the developmental phase, in which major R and D issues are addressed, while the second phase is the design, construction and operation of a pilot plant, called CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell. The third phase is the construction and operation of Demonstration of Fast Reactor Fuel Reprocessing Plant (DFRP) which will provide experience in fast reactor fuel reprocessing with high availability factors and plant throughput. The design, construction and operation of the commercial plant (FRP) for reprocessing of PFBR fuel is the fourth phase, which will provide the requisite confidence for the large scale induction of fast reactors

Decontamination and decommissioning of the SPERT-I Reactor Building at the Idaho National Engineering Laboratory. Final report

International Nuclear Information System (INIS)

Dolenc, M.R.

1986-02-01

This final report documents the decontamination and decommissioning of the SPERT-I Reactor Building. This 20- by 40-ft galvanized steel building was dismantled; and the resultant contaminated sludge, liquid, and carbon steel were disposed of at the Radioactive Waste Management Complex of the Idaho National Engineering Laboratory. This report presents the results of the characterization, decision analysis, planning, and decommissioning of the facility. The total cost of these activities was $139,500. Of this total, $103,500 was required for decommissioning operations. (This latter figure represents a 20% savings over the estimated costs generated during the planning effort.) The objectives of decommissioning this facility were to stabilize the seepage pit area and remove the reactor building. The D and D work was divided into two parts; the seepage pit was decommissioned in 1984, and the reactor building in 1985. The entire area was backfilled with radiologically clean soil, graded, and seeded. Two markers were installed to identify the locations of the pit and reactor building. The only isotopes found in either decommissioning operation were cesium-137 and uranium-235 in very low concentrations. Decommissioning operations of the reactor building were carried out during August 1985. The project generate 297 ft 3 of radioactive waste. No personnel radiation exposure above background was received by D and D workers

31 CFR 500.320 - Domestic bank.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Domestic bank. 500.320 Section 500... § 500.320 Domestic bank. The term domestic bank shall mean any branch or office within the United States of any of the following which is not a national of any designated foreign country: Any bank or trust...

Corrosion of magnesium and some magnesium alloys in gas cooled reactors

International Nuclear Information System (INIS)

Caillat, R.; Darras, R.

1958-01-01

The results of corrosion tests on magnesium and some magnesium alloys (Mg-Zr and Mg-Zr-Zn) in moist air (like G1 reactor) and in CO 2 : (like G2, G3, EDF1 reactors) are reported. The maximum temperature for exposure of magnesium to moist air without any risk of corrosion is 350 deg. C. Indeed, the oxidation rate follows a linear law above 350 deg. C although it reaches a constant level and keeps on very low under 350 deg. C. However, as far as corrosion is concerned this temperature limit can be raised up to 500 deg. C if moist air is very slightly charged with fluorinated compounds. Under pressure of CO 2 , these three materials oxidate much more slowly even if 500 deg. C is reached. The higher is the temperature, the higher is the constant level of the weight increase and the quicker is reached this one. However, Mg-Zr alloy behaves quite better than pure magnesium and especially than Mg-Zr-Zn alloy. (author) [fr

Anaerobic horizontal flow reactor with polyethylene terephthalate as support material

Directory of Open Access Journals (Sweden)

Marcelo Muñoz

2016-06-01

Full Text Available A pilot anaerobic reactor was installed to remove the organic load of wastewater from dairy industry. It uses a bacterial inoculum previously acclimated to the substrate. It was disposed horizontally and filled with pieces of polyethylene terephthalate (PET, from plastic bottles. The reactor was operated at room temperature, during 100 days, in three phases: 1 the reactor was stabilized with volumetric organic load from 0.013 to 0.500 kg/day.m³; 2 the hydraulic retention time was of 1 day and the volumetric organic load of 3 kg/day.m³; 3 the volumetric organic load was incremented from 4 to 6.6 kg/day.m³ and the hydraulic retention time was 1 day. Organic material removal efficiencies was of 85%, and approximately 75% were obtained in the second and third phase, respectively. The Y value was 0.15, indicating that 0.15 kg of biomass were generated by kg of QDO supplied to the reactor. Finally, the biomass generated inside the reactor was analyzed, obtaining a value of 18868 mg/L, which is a higher value than those of conventional systems.

Control rod studies in small and medium sized fast reactors

International Nuclear Information System (INIS)

John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

1988-01-01

Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

Current status of operation and utilization of the Dalat Research Reactor

International Nuclear Information System (INIS)

Dien, Nguyen Nhi

2006-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool-type reactor using the HEU (36% enrichment) WWR-M2 fuel assemblies. It was renovated and upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in the 1st November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analysis, scientific research and training. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for reactor physics and thermal hydraulics experiments. From the first start-up to the end of December 2004, it totaled about 27,253 hrs of operation and the total energy released was about 543 MWd. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 fuel assemblies (FA). The 11 new FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 100 FAs. The second fuel reloading was executed in March 2002. The 4 new FAs were added in the core periphery, at previous beryllium element locations. The working configuration of 104 FAs ensured efficient exploitation of the DNRR at nominal power for about 3000 hrs since March 2002. In order to provide excess reactivity for the reactor operation without the need to discharge high burned FAs, in June 2004, the fuel shuffling of the reactor core was done. 16 FAs with low burn-up from the core periphery were moved toward the core center and 16 FAs with high-burn-up from the core center were moved toward the core periphery. This operation provided additional reactivity of about 0.85 β eff that the current reactor configuration using re-shuffled HEU fuel is expected to allow normal operation until June 2006. In 1999, the request of returning to Russia HEU fuels from foreign

21 CFR 500.92 - Implementation.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Implementation. 500.92 Section 500.92 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS, AND RELATED PRODUCTS GENERAL Regulation of Carcinogenic Compounds Used in Food-Producing Animals...

31 CFR 500.310 - Transfer.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Transfer. 500.310 Section 500.310 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF FOREIGN..., or other fiduciary; the creation or transfer of any lien; the issuance, docketing, filing, or the...

13 CFR 500.212 - Liquidation.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Liquidation. 500.212 Section 500.212 Business Credit and Assistance EMERGENCY OIL AND GAS GUARANTEED LOAN BOARD EMERGENCY OIL AND GAS... proper, consistent with Federal law and regulations. (b) Pursuant to the Guarantee, upon written demand...

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

Science.gov (United States)

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

2014-02-10

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

31 CFR 500.309 - Transactions.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Transactions. 500.309 Section 500.309 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF FOREIGN..., but not by way of limitation: (a) Any payment or transfer to any such designated foreign country or...

Status of fast reactor development in India (April 1996 - March 1997)

International Nuclear Information System (INIS)

Bhoje, S.B.

1998-01-01

India generated 395 TWh of electricity during, 4 April 1996 to March 1997. Oil import bill during the year was $9.3 billion. The operating performance of the thermal power reactors has considerably improved during the year and has enhanced the confidence level in nuclear energy in the government and the public. Construction of 4x220 MWe PHWR is continued at two locations. Start of construction of 2x220 MWe PHWR, 2x500 MWe PHWR and 2x1000 MWe WWER (Russian collaboration) and 500 MWe PFBR have been proposed in the IX Plan (1997- 2002). The 13 party coalition government is discussing the IX plan proposals in the power sector. Operation of FBTR at 10.5 MWt is continued The maximum fuel burnup reached is 32,000 MWd/t without any failure. Targeted burnup is 50,000 MWd/t. Post irradiation examination has been completed on one fuel subassembly taken out at 25,000 MWd/t. The performance of the fuel is very good. Turbine was rolled up to synchronous speed of 3000 rpm several times during the year and operation was found to be smooth. TG synchronisation with grid will be achieved during the reactor operation at 12.5 MWt, with the addition of fuel subassemblies in the core. All the activities related to the revision of conceptual design from 4 loop to 2 loop concept are almost complete for the 500 MWe Prototype Fast Breeder Reactor. The main options for the reactor are sodium coolant, pool type, MOX fuel, 2 primary sodium pumps, 2 secondary loops with 4 SG in each loop. The important design activities carried out during the year are plant dynamic studies, decay heat removal analysis, design of pump to grid plate pipe, scram and LOR parameters, location of secondary sodium pump in the secondary sodium circuit and design of fuel handling machines. Important experimental R and D work carried out during the year were testing of prototype primary sodium pump in water, operation of a large sodium test rig to study the heat and mass transfer in the cover gas, testing of dummy fuel

The Utilization of Dalat nuclear research reactor for education and training purposes

International Nuclear Information System (INIS)

Luong, Ba Vien; Nguyen, Nhi Dien; Le, Vinh Vinh; Nguyen, Xuan Hai

2017-01-01

The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kWt is today the unique one in Vietnam. It was designed for the purposes of radioisotope production, neutron activation analysis, basic and applied researches, and nuclear education and training. With the rising demand in development of human resources for utilization of atomic energy in the country, the DNRR has been playing an important role in the nuclear education and training for students from universities and professionals who are interested in reactor engineering. At present, the Dalat Nuclear Research Institute (DNRI) offers two types of training course utilizing the research reactor: an one-week practical training course is applied for undergraduate students and a two-week training course on reactor engineering is applied for the professionals. This paper presents the reactor facility and experiments performed at the DNRR for education and training purposes. In addition, the co-operation between the DNRI with national and international educational organizations for nuclear human resource development for national and regional demands is also mentioned in the paper. (author)

Current status and ageing management of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Nguyen Nhi Dien [Nuclear Research Institute, Dalat (Viet Nam)

2000-10-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW swimming pool type reactor loaded with the Soviet WWR-M2 fuel elements, moderated and cooled by light water. It was reconstructed and upgraded from the former 250 kW TRIGA Mark-II reactor built in 1963. The first criticality of the renovated reactor was in November 1983 and it has been put in regular operation at nominal power since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs every 4 weeks, for radioisotope production, neutron activation analyses and other research purposes. The remaining time is devoted to maintenance work and to short runs for reactor physics studies as well. From its first start-up to the end of 1998, it totaled about 20,000 hrs of operation at nominal power. After ten years of operation, reactor general inspection and refurbishment were implemented in the 1992-1996 period. In April 1994, refueling work was executed with adding of 11 fresh fuel elements to the reactor core. At present, the reactor has been working with 100-fuel element configuration. Corrosion study has been implemented by visual inspection of the reactor pool tank and some other inside components which remain unchanged from the previous TRIGA reactor. The inspections were carried out with the assistance of some experts from other countries. Some visual inspection results have been obtained and the nature of the electrochemical corrosion and related aspects were little by little identified. In this paper, the operation status of the Dalat reactor is presented, and some activities related to the ageing management of the reactor pool tank and its inside components are also discussed. (author)

Current status and ageing management of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien

2000-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW swimming pool type reactor loaded with the Soviet WWR-M2 fuel elements, moderated and cooled by light water. It was reconstructed and upgraded from the former 250 kW TRIGA Mark-II reactor built in 1963. The first criticality of the renovated reactor was in November 1983 and it has been put in regular operation at nominal power since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs every 4 weeks, for radioisotope production, neutron activation analyses and other research purposes. The remaining time is devoted to maintenance work and to short runs for reactor physics studies as well. From its first start-up to the end of 1998, it totaled about 20,000 hrs of operation at nominal power. After ten years of operation, reactor general inspection and refurbishment were implemented in the 1992-1996 period. In April 1994, refueling work was executed with adding of 11 fresh fuel elements to the reactor core. At present, the reactor has been working with 100-fuel element configuration. Corrosion study has been implemented by visual inspection of the reactor pool tank and some other inside components which remain unchanged from the previous TRIGA reactor. The inspections were carried out with the assistance of some experts from other countries. Some visual inspection results have been obtained and the nature of the electrochemical corrosion and related aspects were little by little identified. In this paper, the operation status of the Dalat reactor is presented, and some activities related to the ageing management of the reactor pool tank and its inside components are also discussed. (author)

44 CFR 19.500 - Employment.

Science.gov (United States)

2010-10-01

... 44 Emergency Management and Assistance 1 2010-10-01 2010-10-01 false Employment. 19.500 Section 19... FEDERAL FINANCIAL ASSISTANCE Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 19.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded...

13 CFR 113.500 - Employment.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Employment. 113.500 Section 113... Discrimination on the Basis of Sex in Employment in Education Programs Or Activities Prohibited § 113.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from participation in, be...

36 CFR 1211.500 - Employment.

Science.gov (United States)

2010-07-01

... 36 Parks, Forests, and Public Property 3 2010-07-01 2010-07-01 false Employment. 1211.500 Section... Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 1211.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from participation in, be...

18 CFR 1317.500 - Employment.

Science.gov (United States)

2010-04-01

... 18 Conservation of Power and Water Resources 2 2010-04-01 2010-04-01 false Employment. 1317.500... Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 1317.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from participation in, be...

38 CFR 23.500 - Employment.

Science.gov (United States)

2010-07-01

... 38 Pensions, Bonuses, and Veterans' Relief 2 2010-07-01 2010-07-01 false Employment. 23.500... Discrimination on the Basis of Sex in Employment in Education Programs or Activities Prohibited § 23.500 Employment. (a) General. (1) No person shall, on the basis of sex, be excluded from participation in, be...

Developments in natural uranium - graphite reactors

International Nuclear Information System (INIS)

Bourgeois, J.

1964-01-01

The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

Ageing of coolant channels in nuclear reactors (PHWRs)

International Nuclear Information System (INIS)

Mitra, T.L.; Chowdhury, M.K.; Gupta, R.K.; Pandarinathan, P.R.; Seth, V.K.

1994-01-01

In PHWRs, ageing of various components takes place due to factors like fast neutron flux, temperature, stress, environment etc. In coolant channel, the most severely affected component due to ageing is pressure tube, though other components like end fitting, calandria tube, garter spring spacer also show ageing to a limited extent. Ageing effects in pressure tube are seen in the form of diametral and axial creep, corrosion, delayed hydrogen cracking and irradiation hardening. In calandria tube and garter spring spacer, creep and hardening are seen though these are not of concern in PHWRs. In end fitting, irradiation embrittlement and abrasion of sealing faces are the areas of concern. Ageing process in these components are the areas of concern. Ageing process in these components are effectively retarded by taking measures like selection of proper material, manufacturing process, control of environmental chemistry, and design modifications. Experience and information gained in various Indian and foreign reactors have been used to improve upon the design in 220 MWe reactors and have formed the basis of design for 500 MWe reactors. (author). 3 refs., 5 figs

N-reactor charge-discharge system analysis

International Nuclear Information System (INIS)

Tokarz, R.D.; Marr, G.D.; Nesbitt, J.F.

1982-09-01

This report documents an analysis of the existing systems in the N-Reactor fuel flow path. It recommends equipment improvements and changes in that path to allow the charge-discharge rates to be increased to 500 tubes per outage without increasing reactor outage time. The estimated program cost of $14 million is projected over an estimated 3-year period. It does not include costs detailed as part of the existing restoration program or any costs that are considered as normal maintenance. The recommendations contained in this report provide a direction and goal for every critical aspect of the fuel flow path. The way in which these recommendations are implemented may greatly affect the schedule and costs. Previous studies by UNC have shown that enhanced fuel element handling has the potential of increasing productivity by 33 days at a cost benefit estimated at $18 million per year. Enhanced fuel handling provides the greatest potential for productivity improvement of any of the areas considered in these studies

Status of sodium cooled fast reactors with closed fuel cycle in India

International Nuclear Information System (INIS)

Raj, B.

2007-01-01

Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

The generation of denatured reactor plutonium by different options of the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Broeders, C.H.M.; Kessler, G. [Inst. for Neutron Physics and Reactor Technology, Research Center Karlsruhe (Germany)

2006-11-15

Denatured (proliferation resistant) reactor plutonium can be generated in a number of different fuel cycle options. First denatured reactor plutonium can be obtained if, instead of low enriched U-235 PWR fuel, re-enriched U-235/U-236 from reprocessed uranium is used (fuel type A). Also the envisaged existing 2,500 t of reactor plutonium (being generated world wide up to the year 2010), mostly stored in intermediate fuel storage facilities at present, could be converted during a transition phase into denatured reactor plutonium by the options fuel type B and D. Denatured reactor plutonium could have the same safeguards standard as present low enriched (<20% U-235) LWR fuel. It could be incinerated by recycling once or twice in PWRs and subsequently by multi-recycling in FRs (CAPRA type or IFRs). Once denatured, such reactor plutonium could remain denatured during multiple recycling. In a PWR, e.g., denatured reactor plutonium could be destroyed at a rate of about 250 kg/GWey. While denatured reactor plutonium could be recycled and incinerated under relieved IAEA safeguards, neptunium would still have to be monitored by the IAEA in future for all cases in which considerable amounts of neptunium are produced. (orig.)

Steady-state plasma and reactor parameters for elliptical cross section tokamaks with very large power ratings

International Nuclear Information System (INIS)

Usher, J.L.; Powell, J.R.

1975-06-01

In previous studies only circular cross section reactor plasmas were considered. The purpose of this research is to examine the effects of elliptical plasma cross sections. Several technological benefits have been determined. Maximum magnetic field strength requirements are 30 to 65 percent less than for 5000 MW (th) reactors and may be as much as 40 percent less than for circular cross section reactors of comparable size. Very large n tau values are found (10 15 to 10 17 sec/cm 3 ), which produce large burn-up fractions (15 to 60 percent). There is relatively little problem with impurity build-up. Long confinement times (60 to 500 seconds) are found. Finally, the elliptical cross section reactors exhibit a major toroidal radius reduction of as large as 30 percent over circular reactors operating at comparable power levels

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. The increase in the price of crude oil in December 1973 changed the economic position of SMPRs so much that even 100-200MW(e) nuclear reactors were considered economic compared with oil-fired plants. The IAEA 1974 market survey showed a potential for 154 units from 150-500MW(e) during 1980-1989 with a total installed capacity of 45000MW(e). This did not generate the desired interest among reactor manufacturers. So far only three European-based manufacturers have shown interest in SMPRs and at present small reactors are being built commercially mostly in India. The reported capital costs of a 215MW(e) Indian CANDU reactor compare favourably with those for European-built reactors. Bangladesh, Jamaica and Kuwait are seriously looking for reactors of 50-150MW(e). The paper analyses the historical background of SMPRs and their commercial application and suggests the following action: (1) A realistic reappraisal of the changed market potential for SMPRs and a critical analysis of the Indian and European figures, possibly carried out by the IAEA; (2) a Special Nuclear Fund be created by contributions from Member States to provide financial support to selected reactor manufacturers willing to make SMPRs commercially available; (3) the proposed Special Nuclear Fund may also provide credit on soft terms to developing countries interested in building SMPRs; (4) the IAEA should expand the scope of its activities and take up the responsibility of collecting and administering such a Special Nuclear Fund. (author)

Fast reactor database. 2006 update

International Nuclear Information System (INIS)

2006-12-01

Liquid metal cooled fast reactors (LMFRs) have been under development for about 50 years. Ten experimental fast reactors and six prototype and commercial size fast reactor plants have been constructed and operated. In many cases, the overall experience with LMFRs has been rather good, with the reactors themselves and also the various components showing remarkable performances, well in accordance with the design expectations. The fast reactor system has also been shown to have very attractive safety characteristics, resulting to a large extent from the fact that the fast reactor is a low pressure system with large thermal inertia and negative power and temperature coefficients. In addition to the LMFRs that have been constructed and operated, more than ten advanced LMFR projects have been developed, and the latest designs are now close to achieving economic competitivity with other reactor types. In the current world economic climate, the introduction of a new nuclear energy system based on the LMFR may not be considered by utilities as a near future option when compared to other potential power plants. However, there is a strong agreement between experts in the nuclear energy field that, for sustainability reasons, long term development of nuclear power as a part of the world's future energy mix will require the fast reactor technology, and that, given the decline in fast reactor development projects, data retrieval and knowledge preservation efforts in this area are of particular importance. This publication contains detailed design data and main operational data on experimental, prototype, demonstration, and commercial size LMFRs. Each LMFR plant is characterized by about 500 parameters: physics, thermohydraulics, thermomechanics, by design and technical data, and by relevant sketches. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors with complete technical information of a total of 37 LMFR

Results of Operation and Utilization of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Le Vinh Vinh; Duong Van Dong; Nguyen Xuan Hai; Pham Ngoc Son; Cao Dong Vu

2014-01-01

The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW was reconstructed and upgraded from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The renovated reactor was put into operation on 20 March 1984. It was designed for the purposes of radioisotope production (RI), neutron activation analysis (NAA), basic and applied researches, and nuclear education and training. During the last 30 years of operation, the DNRR was efficiently utilized for producing many kinds of radioisotopes and radiopharmaceuticals used in nuclear medicine centers and other users in industry, agriculture, hydrology and scientific research; developing a combination of nuclear analysis techniques (INAA, RNAA, PGNAA) and physic-chemical methods for quantitative analysis of about 70 elements and constituents in various samples; carrying out experiments on the reactor horizontal beam tubes for nuclear data measurement, neutron radiography and nuclear structure study; and establishing nuclear training and education programs for human resource development. This paper presents the results of operation and utilization of the DNRR. In addition, some main reactor renovation projects carried out during the last 10 years are also mentioned in the paper. (author)

Activity of Catalase (CAT, ALT and AST in Different Organs of Swiss Albino Mice Treated with Lead Acetate, Vitamin C and Magnesium-L-Threonate

Directory of Open Access Journals (Sweden)

Ilir Nazmi Mazreku

2017-11-01

Full Text Available Introduction: Lead is a natural element with toxic properties and is widespread in the environment. Lead toxicity is associated with generation of reactive oxygen and nitrogen species and consumption of antioxidants elements (vitamin E and C, glutathione, thioredoxin and lipoic acid, melatonin, carotenoids and natural flavonoids in the cell, and unbalancing oxidantsantioxidants levels. Aim: To evaluate the effects of different chemical combinations (lead acetate, Vitamin C and Magnesium-L-threonate on antioxidant enzyme activity (catalase-CAT of liver, kidney, spleen, pancreas and brain, and serum transaminases [Serum Alanine Transaminase (ALT and Serum Aspartate Transaminase (AST]. Materials and Methods: Experimental animals (49 male Mus musculus-swiss albino mice were separated into five different groups. The first group was used as a control, hence the other four groups were treated with sub-lethal doses (90 mg/kg of lead acetate (group 2, lead acetate (90 mg/kg and Vitamin C dose 40mg/kg (group 3, lead acetate (90 mg/kg and Magnesium-Lthreonate dose 100 mg/kg (group 4 and only with MagnesiumL-threonate dose 100 mg/kg (group 5, during the treatment period (40 days. Blood samples were taken from the facial vein and used for transaminase analysis. Organ tissue was collected after euthanizing anaesthetized animals with neck dislocation technique. Results: The results showed that lead acetate treatment has caused significant elevation in the activity of AST (group 2 and 3 and ALT (group 3. Also, CAT activity was significantly (p<0.05 increased in groups treated with lead acetate (liver, pancreas, kidney and brain but not in spleen. Treatment of lead intoxicated groups with Vitamin C and Magnesium L-threonate increased significantly CAT activity in brain. Conclusion: Lead effects by interacting with different molecular systems and increasing enzyme activity (CAT, ALT and AST. Effects on CAT activity of Magnesium-L-threonate and Vitamin C treatment

7 CFR 500.3 - Preservation of property.

Science.gov (United States)

2010-01-01

... 7 Agriculture 6 2010-01-01 2010-01-01 false Preservation of property. 500.3 Section 500.3 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL RESEARCH SERVICE, DEPARTMENT OF AGRICULTURE NATIONAL ARBORETUM Conduct on U.S. National Arboreturm Property § 500.3 Preservation...

10 CFR 600.500 - Purpose and scope.

Science.gov (United States)

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Purpose and scope. 600.500 Section 600.500 Energy DEPARTMENT OF ENERGY (CONTINUED) ASSISTANCE REGULATIONS FINANCIAL ASSISTANCE RULES Eligibility Determination for Certain Financial Assistance Programs-General Statement of Policy § 600.500 Purpose and scope...

European supercritical water cooled reactor

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

2011-01-01

Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

29 CFR 500.80 - Payroll records required.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Payroll records required. 500.80 Section 500.80 Labor... SEASONAL AGRICULTURAL WORKER PROTECTION Worker Protections Wages and Payroll Standards § 500.80 Payroll... agricultural association which employs any migrant or seasonal agricultural worker shall preserve all payroll...

31 CFR 500.578 - Vietnamese property unblocked.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Vietnamese property unblocked. 500.578 Section 500.578 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued..., Authorizations and Statements of Licensing Policy § 500.578 Vietnamese property unblocked. All transactions...

Hydrogen production by enhanced-sorption chemical looping steam reforming of glycerol in moving-bed reactors

International Nuclear Information System (INIS)

Dou, Binlin; Song, Yongchen; Wang, Chao; Chen, Haisheng; Yang, Mingjun; Xu, Yujie

2014-01-01

Highlights: • New approach on continuous high-purity H 2 produced auto-thermally with long time. • Low-cost NiO/NiAl 2 O 4 exhibited high redox performance to H 2 from glycerol. • Oxidation, steam reforming, WSG and CO 2 capture were combined into a reactor. • H 2 purity of above 90% was produced without heating at 1.5–3.0 S/C and 500–600 °C. • Sorbent regeneration and catalyst oxidization achieved simultaneously in a reactor. - Abstract: The continuous high-purity hydrogen production by the enhanced-sorption chemical looping steam reforming of glycerol based on redox reactions integrated with in situ CO 2 removal has been experimentally studied. The process was carried out by a flow of catalyst and sorbent mixture using two moving-bed reactors. Various unit operations including oxidation, steam reforming, water gas shrift reaction and CO 2 removal were combined into a single reactor for hydrogen production in an overall economic and efficient process. The low-cost NiO/NiAl 2 O 4 catalyst efficiently converted glycerol and steam to H 2 by redox reactions and the CO 2 produced in the process was simultaneously removed by CaO sorbent. The best results with an enriched hydrogen product of above 90% in auto-thermal operation for reforming reactor were achieved at initial temperatures of 500–600 °C and ratios of steam to carbon (S/C) of 1.5–3.0. The results indicated also that not all of NiO in the catalyst can be reduced to Ni by the reaction with glycerol, and the reduced Ni can be oxidized to NiO by air at 900 °C. The catalyst oxidization and sorbent regeneration were achieved under the same conditions in air reactor

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

CERN Document Server

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

2015-07-16

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

Paleogeographic features of Zhiguli arch development in the Frasnian century

Energy Technology Data Exchange (ETDEWEB)

Yaroslavtsev, G.A.

1984-01-01

The absence of terrigenous masses of the Devonian on the Pokrovskiy apex of the Zhiguli arch is linked by many researchers with the existence here of an island mass. The unsoundness of this hypothesis is substantiated. The absence of terrigenous masses of the Devonian and the occurrence of offshore carbonate deposits of Voronezh age with discordance on the underlying masses is substantiated by the underwater erosion of underlaying masses and recessive accumulation of carbonate Voronezh formations.

Design of reactor protection systems for HTR plants generating electric power and process heat problems and solutions

International Nuclear Information System (INIS)

Craemer, B.; Dahm, H.; Spillekothen, H.G.

1982-06-01

The design basis of the reactor protection system (RPS) for HTR plants generating process heat and electric power is briefly described and some particularities of process heat plants are indicated. Some particularly important or exacting technical measuring positions for the RPS of a process heat HTR with 500 MWsub(th) power (PNP 500) are described and current R + D work explained. It is demonstrated that a particularly simple RPS can be realized in an HTR with modular design. (author)

29 CFR 500.100 - Vehicle safety obligations.

Science.gov (United States)

2010-07-01

... safety standards. Prima facie evidence that safety standards have been met will be shown by the presence... 29 Labor 3 2010-07-01 2010-07-01 false Vehicle safety obligations. 500.100 Section 500.100 Labor... § 500.100 Vehicle safety obligations. (a) General obligations. Each farm labor contractor, agricultural...

48 CFR 504.500 - Scope of subpart.

Science.gov (United States)

2010-10-01

... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Scope of subpart. 504.500 Section 504.500 Federal Acquisition Regulations System GENERAL SERVICES ADMINISTRATION GENERAL ADMINISTRATIVE MATTERS Electronic Commerce in Contracting 504.500 Scope of subpart. This subpart provides policy and procedure for use of GSA's Electronic...

Study on hydrogen production using the fast breeder reactors (FBR)

International Nuclear Information System (INIS)

Kani, Yoshio

2003-01-01

As the fast breeder reactor (FBR) can effectively convert uranium-238 difficult to carry out nuclear fission at thermal neutron reactors to nuclear fissionable plutonium-239 to use it remarkable upgrading of application on uranium can be performed, to be expected for sustainable energy source. And, by reuse minor actinides of long half-life nuclides in reprocessed high level wasted solutions for fuels of nuclear reactors, reduction of radioactive poison based on high level radioactive wastes was enabled. As high temperature of about 800 centigrade was required on conventional hydrogen production, by new hydrogen production technique even at operation temperature of sodium-cooled FBR it can be enabled. Here were described for new hydrogen production methods applicable to FBR on palladium membrane hydrogen separation method carrying out natural gas/steam modification at reaction temperature of about 500 centigrade, low temperature thermo-chemical method expectable simultaneous simplification of production process, and electrolysis method expected on power load balancing. (G.K.)

Is light water reactor technology sustainable?

International Nuclear Information System (INIS)

Rothwell, G.; Van der Zwaan, B.

2001-01-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor

International Nuclear Information System (INIS)

Urcia, Agustin; Arrieta, Rolando; Baltuano, Oscar; Chan, Renzo; Tincopa, Jean Pierre; Urquizo, Rafael; Rosas, Bernick

2014-01-01

This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

16 CFR 500.4 - Statement of identity.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 1 2010-01-01 2010-01-01 false Statement of identity. 500.4 Section 500.4... PACKAGING AND LABELING ACT § 500.4 Statement of identity. (a) The principal display panel of a consumer commodity shall bear a specification of the identity of the commodity. (b) Such specification of identity...

Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor; Desarrollo de un sistema de vigilancia remota, por medio del monitoreo de parametros claves de seguridad, para un reactor nuclear de investigacion

Energy Technology Data Exchange (ETDEWEB)

Urcia, Agustin; Arrieta, Rolando [Direccion de Produccion, Instituto Peruano de Energia Nuclear, Lima (Peru); Baltuano, Oscar; Chan, Renzo [Direccion de Investigacion y Desarrollo, Instituto Peruano de Energia Nuclear, Lima (Peru); Tincopa, Jean Pierre [Facultad de Ingenieria Electrica y Electronica, Universidad Nacional del Callao, Callao (Peru); Urquizo, Rafael [Facultad de Ingenieria Electronica, Universidad Tecnologica del Peru, Lima (Peru); Rosas, Bernick [Facultad de Ingenieria Electronica, Universidad Nacional de Ingenieria, Lima (Peru)

2014-07-01

This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

Reactor thread-joint metal with corrosion resistant coating material low cycle fatigue

International Nuclear Information System (INIS)

Gorynin, V.I.; Kondratyev, S.Yu.

1991-01-01

The results of test carried out show that the Ni-P plating which was thermally treated in inert medium, provide the dependence of the reactor equipment studs in the high-concentrated medium of leakage for a period of up to 3000 hours. The Al and aluminized platings of the studs made of steel 38 KhN 3 MFA don't provide their corrosion dependence in the reactor medium. Cr plating provides the dependence during 500 hours. The reported test allows to recommend Ni-P plating to depend the studs in the conditions of the effect of the high-concentrated leakage medium, containing KOH, H 3 BO 3 and NaCl. (author)

Removal of Cr(VI) from wastewaters at semi-industrial electrochemical reactors with rotating ring electrodes

International Nuclear Information System (INIS)

Rodriguez R, Miriam G.; Mendoza, Victor; Puebla, Hector; Martinez D, Sergio A.

2009-01-01

In Mexico, most of the electroplating and textile industries are small facilities and release relatively large amounts of hexavalent chromium (Cr(VI)) in surface waters. In this work, the results obtained during the operation of a batch reactor with a capacity of 170 L, and three electrochemical flow reactors-in-series system with a total capacity of 510 L (both using iron rotating ring electrodes to remove Cr(VI) from wastewaters) are presented. The reactors were scaled up from a laboratory reactor to a semi-industrial level, based on the similarity (dynamical, geometrical and electrochemical). An empirical Cr(VI) removal model was validated in batch and continuous reactors at different operating conditions. Cr(VI) concentration of the industrial wastewaters was reduced from about 500 mg/L to values lower than 0.5 mg/L. A very important parameter that affects the process is the pH, which affects the solubility of the Fe(III). Finally, the electrochemical treated wastewater can be reused

Removal of Cr(VI) from wastewaters at semi-industrial electrochemical reactors with rotating ring electrodes

Energy Technology Data Exchange (ETDEWEB)

Rodriguez R, Miriam G. [Depto. Energia, Universidad Autonoma Metropolitana-Azcapotzalco, Av. San Pablo 180, Azcapotzalco, CP 07740, Mexico D.F. (Mexico); Mendoza, Victor [Depto. Electronica, Universidad Autonoma Metropolitana-Azcapotzalco, Av. San Pablo 180, Azcapotzalco, CP 07740, Mexico D.F. (Mexico); Puebla, Hector [Depto. Energia, Universidad Autonoma Metropolitana-Azcapotzalco, Av. San Pablo 180, Azcapotzalco, CP 07740, Mexico D.F. (Mexico); Martinez D, Sergio A. [Depto. Energia, Universidad Autonoma Metropolitana-Azcapotzalco, Av. San Pablo 180, Azcapotzalco, CP 07740, Mexico D.F. (Mexico)], E-mail: samd@correo.azc.uam.mx

2009-04-30

In Mexico, most of the electroplating and textile industries are small facilities and release relatively large amounts of hexavalent chromium (Cr(VI)) in surface waters. In this work, the results obtained during the operation of a batch reactor with a capacity of 170 L, and three electrochemical flow reactors-in-series system with a total capacity of 510 L (both using iron rotating ring electrodes to remove Cr(VI) from wastewaters) are presented. The reactors were scaled up from a laboratory reactor to a semi-industrial level, based on the similarity (dynamical, geometrical and electrochemical). An empirical Cr(VI) removal model was validated in batch and continuous reactors at different operating conditions. Cr(VI) concentration of the industrial wastewaters was reduced from about 500 mg/L to values lower than 0.5 mg/L. A very important parameter that affects the process is the pH, which affects the solubility of the Fe(III). Finally, the electrochemical treated wastewater can be reused.

Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

1995-09-01

Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

Comparison of performance indicators of different types of reactors based on ISOE database

International Nuclear Information System (INIS)

Janzekovic, H.; Krizman, M.

2005-01-01

The optimisation of the operation of a nuclear power plant (NPP) is a challenging issue due to the fact that besides general management issues, a risk associated to nuclear facilities should be included. In order to optimise the radiation protection programmes in around 440 reactors in operation with more than 500 000 monitored workers each year, the international exchange of performance indicators (PI) related to radiation protection issues seems to be essential. Those indicators are a function of a type of a reactor as well as the age and the quality of the management of the reactor. in general three main types of radiation protection PI could be recognised. These are: occupational exposure of workers, public exposure and management of PI related to radioactive waste. The occupational exposure could be efficiently studied using ISOC database. The dependence of occupational exposure on different types of reactors, e.g. PWR, BWR, are given, analysed and compared. (authors)

Strain components of nuclear-reactor-type concretes during first heat cycle

International Nuclear Information System (INIS)

Khoury, G.A.

1995-01-01

Strains of three advanced-gas-cooled-reactor-type nuclear reactor concretes were measured during the first heat cycle and their relative thermal stability determined. It was possible to isolate for the first time the shrinkage component for the period during heating. Predictions of the residual strains for the loaded specimens can be made by simple superposition of creep and shrinkage components up to a certain critical temperature, which for basalt concrete is about 500 C and for limestone concrete is about 200-300 C. Above the critical temperature, an expansive ''cracking'' strain component is present. It is shown that the strain behaviour of concrete provides a sensitive indication of its thermal stability during heating and subsequent cooling. (orig.)

12 CFR 411.500 - Secretary of Defense.

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 4 2010-01-01 2010-01-01 false Secretary of Defense. 411.500 Section 411.500 Banks and Banking EXPORT-IMPORT BANK OF THE UNITED STATES NEW RESTRICTIONS ON LOBBYING Exemptions § 411.500 Secretary of Defense. (a) The Secretary of Defense may exempt, on a case-by-case basis, a covered...

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

Energy Technology Data Exchange (ETDEWEB)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki [Nuclear Research Group, FMIPA, Bandung Institute of Technology Jl. Ganesha 10, Bandung 40132 (Indonesia); Miura, Ryosuke; Takaki, Naoyuki [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, H. [Emerritus Prof. of Research Laboratory for Nuclear Reactors, Tokyo Inst. of Technology (Japan)

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

78 FR 75511 - Special Conditions: Bombardier Inc., Models BD-500-1A10 and BD-500-1A11 Series Airplanes...

Science.gov (United States)

2013-12-12

... Inc., Models BD-500-1A10 and BD- 500-1A11 Series Airplanes; Electronic Flight Control System: Control... Inc. Models BD-500-1A10 and BD-500-1A11 series airplanes. These airplanes will have a novel or unusual... comments, data, or views. The most helpful comments reference a specific portion of the special conditions...

77 FR 69568 - Special Conditions: Bombardier Aerospace, Model BD-500-1A10 and BD-500-1A11 Airplanes; Sidestick...

Science.gov (United States)

2012-11-20

... Aerospace, Model BD-500-1A10 and BD-500-1A11 Airplanes; Sidestick Controllers AGENCY: Federal Aviation... conditions for the Bombardier Aerospace Model BD-500-1A10 and BD-500-1A11 airplanes. These airplanes will... on the comments we receive. Background On December 10, 2009, Bombardier Aerospace applied for a type...

Processing test of an upgraded mechanical design for PERMCAT reactor

International Nuclear Information System (INIS)

Borgognoni, Fabio; Demange, David; Doerr, Lothar; Tosti, Silvano; Welte, Stefan

2010-01-01

The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H 2 O and D 2 .

9 CFR 319.500 - Meat pies.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Meat pies. 319.500 Section 319.500... ORGANIZATION AND TERMINOLOGY; MANDATORY MEAT AND POULTRY PRODUCTS INSPECTION AND VOLUNTARY INSPECTION AND CERTIFICATION DEFINITIONS AND STANDARDS OF IDENTITY OR COMPOSITION Meat Food Entree Products, Pies, and...

41 CFR 101-4.500 - Employment.

Science.gov (United States)

2010-07-01

... 41 Public Contracts and Property Management 2 2010-07-01 2010-07-01 true Employment. 101-4.500... Employment in Education Programs or Activities Prohibited § 101-4.500 Employment. (a) General. (1) No person... subjected to discrimination in employment, or recruitment, consideration, or selection therefor, whether...

Safety problems of nuclear power plants with reactors of new generation; Voprosy bezopasnosti v proehktakh AEhS novogo pokoleniya s WWER

Energy Technology Data Exchange (ETDEWEB)

Fedorov, V; Rogov, M; Biryukov, G [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

1996-12-31

Modernization schemes for safety enhancement of WWER-1000 reactors are proposed. In the case of WWER-1000/V-392 it is based on introduction of additional safety systems and overall design improvement. For WWER-1100 (1000-1100 MW) the safety is enhanced by passive systems built from two-stage heat exchangers. For WWER-500/600 the use of passive safety system is extended to emergency cooling of the active zone and removal of the residual heat emissions from the reactor. The technical characteristics of the three reactors are compared. 3 figs., 1 tab.

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

A Gas-Cooled Reactor Surface Power System

Energy Technology Data Exchange (ETDEWEB)

Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

1998-11-09

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

Design of shutdown system no.2 liquid poison injection system for 500 MWe PHWR

International Nuclear Information System (INIS)

Bhatnagar, S.; Balasubrahmanian, A.K.; Pillai, A.V.

1997-01-01

Defence in depth and two group system concepts form the basic design philosophy for the shutdown systems. There are two independent, diverse and fast acting shutdown systems provided for the 500 MWe PHWR. The design is based on fail-safe principle, sufficient component redundancy and on-line testing. Liquid poison injection system, as shutdown system 2, is newly developed for the 500 MWe PHWRs. The system operates by rapidly injecting gadolinium nitrate solution into bulk moderator using stored helium pressure thereby inserting negative reactivity. A high pressure helium supply tank which provides the energy for system actuation, is connected, through an array of fast acting valves in series-parallel arrangement, to the individual poison tanks storing gadolinium nitrate solution. The valves, belonging to three different channels of reactor Protection System 2, are the only active components in the system. The valves are fail safe and are periodically tested on-line without actually firing the system. The system comprising of in-core assemblies and the external process system has been engineered. Experimental work is being carried out by BARC for design validation and data generation. This paper describes the conceptual development, design basis, design parameters and detailed engineering of the system. (author)

31 CFR 500.314 - Banking institution.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Banking institution. 500.314 Section 500.314 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... or incidentally in the business of banking, of granting or transferring credits, or of purchasing or...

Concept of a nuclear powered submersible research vessel and a compact reactor

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa; Nishimura, Hajime; Tokunaga, Sango

2001-07-01

A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

Design of a reactor system for the synthesis of titanium diboride

International Nuclear Information System (INIS)

Tsui, M.E.; Epstein, H.A.

1981-10-01

TiB 2 , a hard, refractory material, is difficult to produce at a purity required for many potential uses. In this study, a laboratory-scale reactor system was designed to produce 4 g/h of very pure TiB 2 powder from a homogeneous-nucleation gas-phase reaction. The system operates at temperatures up to 1700 K, pressures from 1 torr to 1 atm, and incorporates a novel flame-reactor concept in which the heat of reaction for powder formation is provided by a H 2 -Cl 2 flame. The powder is produced in an alumina reactor 3-3/8-in. -ID x 55-in. long, with a feed preheater, and is collected in low-pressure traps. The system is fully instrumented for study of reaction kinetics and powder morphology. System startup, operating, shutdown, and safety procedures as well as a proposed experimental plan are included. The estimated construction cost of the system is $24,500

Feasibility study for Tehran Research Reactor power upgrading

Energy Technology Data Exchange (ETDEWEB)

Farhadi, Kazem [Nuclear Research Center, Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)], E-mail: kfarhadi@aeoi.org.ir; Khakshournia, Samad [Nuclear Research Center, Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

2008-07-15

The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW{sub th} to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5-11 MW) and different core coolant flow rates (500-921 m{sup 3}/h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW{sub th} with the total power peaking factor maintained at less than or equal to 3.0.

Feasibility study for Tehran Research Reactor power upgrading

International Nuclear Information System (INIS)

Farhadi, Kazem; Khakshournia, Samad

2008-01-01

The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW th to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5-11 MW) and different core coolant flow rates (500-921 m 3 /h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW th with the total power peaking factor maintained at less than or equal to 3.0

Primary system thermal hydraulics of future Indian fast reactors

Energy Technology Data Exchange (ETDEWEB)

Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

2015-12-01

Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

2003-01-01

The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

Removal of the Materials Test Reactor overhead working reservoir

International Nuclear Information System (INIS)

Lunis, B.C.

1975-10-01

Salient features of the removal of an excessed contaminated facility, the Materials Test Reactor (MTR) overhead working reservoir (OWR) from the Test Reactor Area to the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory are described. The 125-ton OWR was an overhead 160,000-gallon-capacity tank approximately 193 feet high which supplied cooling water to the MTR. Radiation at ground level beneath the tank was 5 mR/hr and approximately 600 mR/hr at the exterior surface of the tank. Sources ranging from 3 R/hr to in excess of 500 R/hr exist within the tank. The tank interior is contaminated with uranium, plutonium, and miscellaneous fission products. The OWR was lowered to ground level with the use of explosive cutters. Dismantling, decontamination, and disposal were performed by Aerojet Nuclear Company maintenance forces

Is light water reactor technology sustainable?

Energy Technology Data Exchange (ETDEWEB)

Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

2001-07-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

Safety aspects of using gadolinium as burnable poison in pressurized water reactors

International Nuclear Information System (INIS)

Vandenberg, C.; Bonet, H.; Charlier, A.; Motte, F.

1979-01-01

Within the framework of an experimental program on the behavior of gadolinium in light water reactors (LWRs), the BR3 power plant, a small 11-MW(electric) pressurized water reactor, was operated successfully with a core containing 5% Gd 2 O 3 -UO 2 rods. The core reached an average burnup increase of 22,000 MWd/tM, corresponding to 500 effective full-power days in a single cycle. These results were used to extrapolate the consequences on safety of extending such a control policy to large LWRs. In this context, the following factors were investigated: impact on the design, reactivity control and core behavior operated with lower and more constant boric acid concentration, environmental impact, fuel handling, etc

Production of bio-oil from underutilized forest biomass using an auger reactor

Science.gov (United States)

H. Ravindran; S. Thangalzhy-Gopakumar; S. Adhikari; O. Fasina; M. Tu; B. Via; E. Carter; S. Taylor

2015-01-01

Conversion of underutilized forest biomass to bio-oil could be a niche market for energy production. In this work, bio-oil was produced from underutilized forest biomass at selected temperatures between 425–500°C using an auger reactor. Physical properties of bio-oil, such as pH, density, heating value, ash, and water, were analyzed and compared with an ASTM standard...

15 CFR 8a.500 - Employment.

Science.gov (United States)

2010-01-01

... 15 Commerce and Foreign Trade 1 2010-01-01 2010-01-01 false Employment. 8a.500 Section 8a.500 Commerce and Foreign Trade Office of the Secretary of Commerce NONDISCRIMINATION ON THE BASIS OF SEX IN...-sponsored activities, including social or recreational programs; and (10) Any other term, condition, or...

31 CFR 500.307 - Unblocked national.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Unblocked national. 500.307 Section 500.307 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF...” shall not be deemed to suspend in any way the requirements of any section of this chapter relating to...

Calculation of hydrostatic radial bearing for main circulating pump of 500 BIKS type

International Nuclear Information System (INIS)

Hnatek, T.; Sojka, P.

1978-01-01

Computer calculations of the radial hydrostatic bearing were performed for the main circulating pump of the 500 BIKS type designed for WWER reactors. The calculations were based on the Reynolds equation of thin layer hydrodynamic pressure in turbulent flow. Relations were derived for orifice reducer flow. In contrast to previous calculations conducted for laminar flow, the results are more accurate because the nature of bearing lubrication evidently is turbulent. The required loading of 21,700 N in normal pump operation is fully compensated at a full eccentricity of 0.77. Operating tests of the pump also confirmed that the actual radial forces on the rotor did not attain the desired loading. On the other hand, thanks to the bearing brass design, the bearing is capable of short-time operation with limit eccentricity, ie., at start, in deceleration and in emergency conditions. (Z.M.)

Power start up of the Dalat nuclear research reactor; Khoi dong nang luong lo phan ung hat nhan Da Lat

Energy Technology Data Exchange (ETDEWEB)

Hien, Pham Duy; Huy, Ngo Quang; Long, Vu Hai; Mai, Tran Khanh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

After accomplishing the physical start-up of the reactor, the power start-up was carried out in February 1984. The power of the reactor has reached: 10 KW on 6/2/1984, 100 KW on 7/2/1984, 200 KW and 300 KW on 8/2/1984; 400 KW and nominal power 500 KW on 9/2/1984. The reactivity temperature coefficient and the xenon poisoning were determined. 3 figs., 12 tabs.

46 CFR 16.500 - Management Information System requirements.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Management Information System requirements. 16.500 Section 16.500 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY MERCHANT MARINE OFFICERS AND SEAMEN CHEMICAL TESTING Management Information System § 16.500 Management Information System requirements. (a...

14 CFR 1203.500 - Use of derivative classification.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 5 2010-01-01 2010-01-01 false Use of derivative classification. 1203.500 Section 1203.500 Aeronautics and Space NATIONAL AERONAUTICS AND SPACE ADMINISTRATION INFORMATION SECURITY PROGRAM Derivative Classification § 1203.500 Use of derivative classification. The application of...

Study of the $B^0 \\rightarrow K^{\\ast 0}e^+ e^-$ decay with the LHCb detector and development of a novel concept of PID detector: the Focusing DIRC

CERN Document Server

AUTHOR|(CDS)2088184; Arnaud, Nicolas

Flavour-changing neutral current processes of the type $b \\to s\\gamma$ are forbidden at the tree level in the Standard Model (SM) and occur at leading order through radiative loop diagrams. Therefore, they are sensitive to new physics (NP), which may contribute with competing diagrams. Furthermore, the chirality of the weak interaction in the SM implies that the photon emitted has left-handed polarisation. However, a whole class of NP theories do not share this SM feature and may manifest unambiguously as a right-handed contribution to the polarisation. This thesis presents the first study of the ${b}{s\\gamma}$ photon polarisation through an angular analysis of the $B^0 \\rightarrow K^{\\ast 0}e^+ e^-$ channel. Even though $B^0 \\rightarrow K^{\\ast 0}e^+ e^-$ is not a radiative $b\\to s$ transition, the contribution from a virtual photon coupling to the lepton pair dominates in the low-$q^2$ region. Furthermore, the channel with electrons rather than muons allows to better isolate the virtual photon contribution ...

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

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

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

Summary of current NEACRP views on fast reactor breeding assessment

International Nuclear Information System (INIS)

Barre, J.

1980-01-01

The global breeding gain (GBC), which may be divided into internal breeding gain (IBG) and external breeding gain (EGB), is dealt with for mixed oxide fuelled LMFBR. Relative contributions of core and blankets to GBG are indicated for three power levels (250, 500 and 1200 MWe). Reactor physics studies are performed to reduce uncertainties on GBC. The studies are of three types, depending on countries. The mock-up approach consists of measuring on one critical assembly, typical of the considered power reactor, the GBG at one time of life of the plant, usually the beginning of life configuration (absorbers in) and trying to obtain bias factors. Parametric analysis of the neutron balance and data adjustment in which global parameters of the neutron balance are measured systematically is the approach followed in the UK and France for all configurations of the reactor, especially for integral parameters related to GBG. Analysis of irradiated fuels involves the measurements of the variation of fuel isotopic compositions versus burn-up with two main goals: accurate measurement of captive ratios and global check of the GBG calculation. (UK)

PERKEMBANGAN BIOFILM NITRIFIKASI DI FIXED BED REACTOR PADA SALINITAS TINGGI

Directory of Open Access Journals (Sweden)

Sudarno

2012-03-01

Full Text Available Development of nitrification biomass that is growing attached on carried material was examined by measuring its ammonium or nitrit oxidation rates. Porous ceramic rings (36 pieces were put into the fixed bed reactor (FBR . The fixed bed reactor that was operated continuously for more than 500 day was continued to be operated at a HRT of 1 day, a DO of above 5 mg L-1 and pH of 8. Ammonia concentration in the feeding was 50 mg NH4+-N L-1. At days 1, 5, 12, 20, 33 and 50, six porous ceramic rings were taken out and then ammonia and nitrite removal rate by biofilm in the ceramic rings was separately measured. The measurement of rates was done in small cylindrical glass reactors with initial concentration of ammonia and nitrite was 10 mg N L-1. Until 50 days of incubation AORs were always higher than NORs. Additionally, ammonia oxidizers attach or grow faster in the porous ceramic material than nitrite oxidizers.

29 CFR 500.78 - Information in foreign language.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Information in foreign language. 500.78 Section 500.78... § 500.78 Information in foreign language. Each farm labor contractor, agricultural employer and... English or, as necessary and reasonable, in Spanish or another language common to migrant or seasonal...

Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

Energy Technology Data Exchange (ETDEWEB)

Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

1974-07-01

Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

Trade as an indicator of social and economic development

Directory of Open Access Journals (Sweden)

N. A. Serebryakova

2018-01-01

Full Text Available In modern conditions of trade in the Russian Federation became the most important type of entrepreneurship. It is, in a number of objective and subjective reasons, is the most rapidly developing sector of the national economy, affecting the interests of all subjects of market relations: population, manufacturers of commercial products, government and trade. Currently in the internal trade of the Russian Federation there have been significant changes occurring under the influence of growing tensions on the international market, in the economy of our country and within the trading industry. But despite the deteriorating economic situation in the world and strained relations between the Russian Federation and its European and American partners, the latest statistics indicate the translational dynamics of retail trade turnover in the whole country, and in the Voronezh region. Retail trade turnover is among the most important indicators of economic and social development of the Voronezh region and the country as a whole. Its structure and volume characterize the level of consumption of goods population, the increase or decrease of welfare of the people. Through retail sales is a constant influence on the development of the volume and structure of production of consumer goods. This article assessed the relationship between the economic development of the Voronezh region as one of average of region of our country, its trade and standard of living of Voronezh.

21 CFR 500.46 - Hexachlorophene in animal drugs.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Hexachlorophene in animal drugs. 500.46 Section 500.46 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS, AND RELATED PRODUCTS GENERAL Specific Administrative Rulings and Decisions § 500.46...

9 CFR 381.500 - Exemption from nutrition labeling.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Exemption from nutrition labeling. 381.500 Section 381.500 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE, DEPARTMENT OF... business is any single-plant facility or multi-plant company/firm that employs 500 or fewer people and...

Processing test of an upgraded mechanical design for PERMCAT reactor

Energy Technology Data Exchange (ETDEWEB)

Borgognoni, Fabio, E-mail: fabio.borgognoni@enea.i [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Demange, David; Doerr, Lothar [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany); Tosti, Silvano [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Welte, Stefan [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany)

2010-12-15

The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H{sub 2}O and D{sub 2}.

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Automation of the radiation protection monitoring system in the RP-10 reactor

International Nuclear Information System (INIS)

Anaya G, Olgger; Castillo Y, Walter; Ovalle S, Edgar

2002-01-01

During the reactor operation, it is necessary to carry out the radiological control in the different places of the reactor, in periodic form and to take a registration of these values. For it the radioprotection official, makes every certain periods, settled down in the procedures, to verify and to carry out the registration of those values in manual form of each one of the radiation monitors. For this reason it was carried out the design and implementation of an automatic monitoring system of radioprotection in the reactor. In the development it has been considered the installation of a acquisition data system for 27 radiation gamma monitors of the type Geiger Mueller, installed inside the different places of the reactor and in the laboratories where they are manipulated radioactive material, using as hardware the FieldPoint for the possessing and digitalization of the signs which are correspondents using the communication protocol RS-232 to a PC in which has settled a program in graphic environment that has been developed using the tools of the programming software LabWindows/CVI. Then, these same signs are sent 'on line' to another PC that is in the Emergency Center of Coordination to 500 m of the reactor, by means of a system of radiofrequency communication. (author)

The challenge of introducing high-temperature reactor plants onto the international power plant market

International Nuclear Information System (INIS)

Bogen, J.; Stoelzl, D.

1987-01-01

Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

The challenge of introducing high-temperature-reactor plants onto the international power plant market

International Nuclear Information System (INIS)

Bogen, J.; Stoelzl, D.

1988-01-01

Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

500 French verbs for dummies

CERN Document Server

Erotopoulos

2013-01-01

Vexed by French verbs? Fear no more! In 500 French Verbs For Dummies, beginning French language learners can find a quick reference for verbs in the basic present tenses. More advanced French speakers can utilize this book to learn more complex verb tenses and conjugations as well as advanced verbs with irregular endings. One page for each of the 500 most commonly used verbs in the French language -alphabetically arranged and numbered for easy referenceSpecial designation of the 50 most essential French verbsA summary of basic French grammar that incl

29 CFR 500.143 - Civil money penalty assessment.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Civil money penalty assessment. 500.143 Section 500.143... MIGRANT AND SEASONAL AGRICULTURAL WORKER PROTECTION Enforcement § 500.143 Civil money penalty assessment. (a) A civil money penalty may be assessed for each violation of the Act or these regulations. (b) In...

Fortune 500 Corporate Headquarters

Data.gov (United States)

Department of Homeland Security — Large Corporate Headquarters in the United States This database is composed of 'an annual list of the 500 largest industrial corporations in the U.S., published by...

Supply of appropriate nuclear technology for the developing world: small power reactors for electricity generation

International Nuclear Information System (INIS)

Heising-Goodman, C.D.

1981-01-01

This paper reviews the supply of small nuclear power plants (200 to 500 MWe electrical generating capacity) available on today's market, including the pre-fabricated designs of the United Kingdom's Rolls Royce Ltd and the French Alsthom-Atlantique Company. Also, the Russian VVER-440 conventionally built light-water reactor design is reviewed, including information on the Soviet Union's plans for expansion of its reactor-building capacity. A section of the paper also explores the characteristics of LDC electricity grids, reviewing methods available for incorporating larger plants into smaller grids as the Israelis are planning. Future trends in reactor supply and effects on proliferation rates are also discussed, reviewing the potential of the Indian 220 MWe pressurised heavy-water reactor, South Korean and Jananese potential for reactor exports in the Far East, and the Argentine-Brazilian nuclear programme in Latin America. This study suggests that small reactor designs for electrical power production and other applications, such as seawater desalination, can be made economical relative to diesel technology if traditional scaling laws can be altered by adopting and standardising a pre-fabricated nuclear power plant design. Also, economy can be gained if sufficient attention is concentrated on the design, construction and operating experience of suitably sized conventionally built reactor systems. (author)

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

2000-01-01

U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

Cutting technique for reactor internals by laser beam

International Nuclear Information System (INIS)

Matsumoto, O.; Sugihara, M.; Matsuda, K.; Miya, K.

1990-01-01

At present in Japan the verification tests on the commercial nuclear power reactor decommissioning technology are being conducted as the project of The Ministry of International Trade and Industry by Nuclear Power Engineering Test Center. This paper summarizes the interim results of the verification test for the reactor core internals decommissioning technology, which is being conducted from 1986 as a theme of the above project. All core internals to be studied here are made of stainless steel, and the maximum wall thickness is about 500mm (the maximum one to be cut is about 300mm) for the PWR's, and about 100mm for the BWR's. Though the plasma cutting, arc saw cutting method, etc. have been studied u p to now as the cutting technology for decommissioning these core internals, the authors are carrying out the development and verification test of the cutting technology with the laser beam, which is expected to increase its power in future and can be applied to various materials

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

78 FR 11089 - Special Conditions: Bombardier Aerospace, Model BD-500-1A10 and BD-500-1A11 Airplanes; Sidestick...

Science.gov (United States)

2013-02-15

...; Special Conditions No. 25-477-SC] Special Conditions: Bombardier Aerospace, Model BD-500-1A10 and BD-500... special conditions. SUMMARY: These special conditions are issued for the Bombardier Aerospace Model BD-500...: Background On December 10, 2009, Bombardier Aerospace applied for a type certificate for their new Model BD...

Phenol degradation in an anaerobic fluidized bed reactor packed with low density support materials

Directory of Open Access Journals (Sweden)

G. P. Sancinetti

2012-03-01

Full Text Available The objective of this research was to study phenol degradation in anaerobic fluidized bed reactors (AFBR packed with polymeric particulate supports (polystyrene - PS, polyethylene terephthalate - PET, and polyvinyl chloride - PVC. The reactors were operated with a hydraulic retention time (HRT of 24 h. The influent phenol concentration in the AFBR varied from 100 to 400 mg L-1, resulting in phenol removal efficiencies of ~100%. The formation of extracellular polymeric substances yielded better results with the PVC particles; however, deformations in these particles proved detrimental to reactor operation. PS was found to be the best support for biomass attachment in an AFBR for phenol removal. The AFBR loaded with PS was operated to analyze the performance and stability for phenol removal at feed concentrations ranging from 50 to 500 mg L-1. The phenol removal efficiency ranged from 90-100%.

Characteristics of outage radiation fields around various reactor components

International Nuclear Information System (INIS)

Verzilov, Y.; Husain, A.; Corbin, G.

2008-01-01

Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

Training courses for the staff of the nuclear power station KRSKO conducted at the TRIGA reactor center in Ljubljana

International Nuclear Information System (INIS)

Pregl, G.; Najzer, M.

1976-01-01

The training program for the Nuclear Power Station Krsko was divided into two modules: fundamentals of nuclear engineering and specialized training according to duties that candidates are supposed to take at the power station. Basic training was organized at the TRIGA Reactor Center in Ljubljana in two different versions. The first version intended for plant operators and all engineers lasted for six months and included about 500 hours of classroom lessons and seminars and 31 laboratory experiments. The educational program was conventional. The following topics were covered: nuclear and atomic physics, reactor theory, reactor dynamics, reactor instrumentation and control, heat transfer in nuclear power plants, nuclear power plant systems, reactor materials, reactor safety, and radiation protection. Until now, two groups, consisting of 37 candidates altogether, have attended this basic course. Plans have been made to conduct two additional courses of about 20 students each for technicians other than operators. The program of this second version will be reduced, with the emphasis on reactor core physics and radiation protection. Classroom lessons will be strongly supported by laboratory experiments. (author)

Development of in-core measurements in the reactor KS-150

International Nuclear Information System (INIS)

Rana, S.B.

1977-01-01

Mapping of the neutron flux density distribution and of the neutron fluence distribution in the KS-150 reactor core was carried out using an in-core measuring system. The system allows the in-service monitoring of important operating properties of the reactor core and fuel elements and consists of a mapping fuel element assembly with built-in SPN detectors, of transmission paths and a computer facility. The measurement of the neutron flux, neutron fluence and temperature fields in the reactor core was carried out during the power start-up of the reactor using self-powered DPZ-1 detectors. The obtained data are given and the axial distribution of neutron flux is graphically represented for different values of burnup at the same configuration of regulating rods, as is the axial distribution of neutron fluence for different configurations of the regulating rods during operation, and the in-service neutron fluence distribution. The maximal fuel temperature of 500.2 degC was found at a distance of 291.2 cm from the upper boundary of the reactor core, at a neutron flux of 1.46x10 14 n/cm 2 s. In comparison with other methods, this method proved easy and quick, the results reliable, reactivity perturbance negligible and the fuel element cost increase a negligible 4%. Neutron flux mapping using in-core self-powered detectors will be performed on a wider scale. (J.P./J.O.)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

500 Cities: City Boundaries

Data.gov (United States)

U.S. Department of Health & Human Services — This city boundary shapefile was extracted from Esri Data and Maps for ArcGIS 2014 - U.S. Populated Place Areas. This shapefile can be joined to 500 Cities...

29 CFR 780.305 - 500 man-day provision.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false 500 man-day provision. 780.305 Section 780.305 Labor...) Statutory Provisions § 780.305 500 man-day provision. (a) Section 3(u) of the Act defines man-day to mean “any day during which an employee performs agricultural labor for not less than 1 hour.” 500 man-days...

Manufacturing requirements of reactor assembly components for PFBR (Paper No. 041)

International Nuclear Information System (INIS)

Murty, C.G.K.; Bhoje, S.B.

1987-02-01

This paper enumerates the requirements of 500 MWe Prototype Fast Breeder Reactor (PFBR) components and considering the present state of art of Indian industry an analysis is made on the challenges to be faced in manufacture highlighting the areas needing development. The large sizes and weights of the components coupled with the limitations on shop facilities and ODC transport, demand part of the fabrication to be done at shop and balance assembly work as well as certain assembly machining operations to be done at site work shop. The stringent geometrical tolerances coupled with extensive destructive and non-destructive examinations call for balanced and low heat input welding techniques and special inspection equipment like electronic co-ordinate determination system. The present paper deals with the specific manufacturing problems of the main reactor components. (author)

Avaliação hematológica e dosagem bioquímica de ALT, AST e creatinina em elefante-marinho-do-sul, Mirounga leonina (linnaeus, 1758, encontrado no litoral de Salvador, Bahia

Directory of Open Access Journals (Sweden)

Bruno Lopes Bastos

2006-02-01

Full Text Available Since 1999 the Aquatic Mammals Rescue Center - AMRC has been working in the rescue and rehabilitation of stranded cetaceans and pinnipeds on the coast of Bahia, Brazil. This paper presents and analyses the blood cells count and clinical chemistry of alanine aminotransferase (ALT, aspartate aminotransferase (AST and creatinine of a southern elephant seal, Mirounga leonina (LINNAEUS, 1758, found on February the 11th at Barra Beach, Salvador, BA. The specimen was an orphan male calf, with 137cm of length and estimated weight of 49kg. It presented bad nutritional conditions and a shark bite on the right shoulder area. Clinical management was performed for 56 days, anthelmintic Febendazole was utilized, and the bite was treated with iodined alcohol, Nitrofurazone solution and Kethanserin, simultaneously with Enrofloxacin 10%, Potenay®, Vitamin B Complex and Benerva®. On the 16th the animal presented a right unilateral conjuntivitis, treated with Cloranphenicol oftalmic pomade until the end of its stay in the captive. During this period a total of six blood samples were collected, three for total blood counts and the others for the biochemistry determination of ALT, AST and creatinine. According to the haematological analysis the seal developed an anaemia which was classified as microcytic and normochromic. Lymphopenia, eosinopenia and monocytopenia were also observed, possibly due to its handling and stress conditions. The clinical chemistry presented low values for AST and creatinine, although this did not represent the existence of any pathologic context or disease with clinical significance.

Analysis of minor actinides transmutation for a Molten Salt Fast Reactor

International Nuclear Information System (INIS)

Yu, Chenggang; Li, Xiaoxiao; Cai, Xiangzhou; Zou, Chunyan; Ma, Yuwen; Han, Jianlong; Chen, Jingen

2015-01-01

Highlights: • The transmutation of MA in a 500 MWth MSFR is analyzed. • A larger MA loading can enhance the MA transmutation and deepen the burnup. • The MA transmutation efficiency can reach 95%. • The FTC can satisfy the safe operating requirement during the entire operating. - Abstract: As one of the six candidate reactors chosen by the Generation IV International Forum (GIF), Molten Salt Fast Reactor (MSFR) has many outstanding advantages and features for advanced nuclear fuel utilization. Effective transmutation of minor actinides (MA) could be attained in this kind of fast reactor, which is of importance in the future closed nuclear fuel cycle scenario. In this work, we attempt to study the MA transmutation capability in a MSFR with power of 500 MWth by analyzing the neutronics characteristics for different MA loadings. The calculated results show that MA loading plays an important role in the reactivity evolution of the MSFR. A larger MA loading is favorable to improving the MA transmutation performance and simultaneously to reducing the fissile consumption. When MA = 18.17 mol%, the transmutation fraction can achieve to about 95% on iso-breeding. We also find that although the fuel temperature coefficient (FTC) decreases with the increasing MA loading, it is still negative enough to keep the safety of the MSFR during the whole operation time. The MA contribution to the effective delayed neutron fraction (EDNF) and the intensity of spontaneous fission neutron (ISFN) are also analyzed. Also MA loading can affect the EDNF during the operation and the ISFN of the MSFR is dominated by 244 Cm. Finally, we analyze the effect of the core power on MA transmutation capability. The result shows that for all the operating powers the depletion ratio of MA to HN increases with time and reaches a maximum value. And additional MA should be fed into the fuel salt before the MA depletion ratio reaches the peak value to improve its transmutation capability. The net

Development of the digitalized automatic seismic trip system for nuclear AR power plants using the systems engineering approach

International Nuclear Information System (INIS)

Jung, Jae Cheon

2014-01-01

The automatic seismic trip system (ASTS) continuously monitors PGA (peak ground acceleration) from the seismic wave, and automatically generates a trip signal. This work presents how the system can be designed by using a systems engineering approach under the given regulatory criteria. Overall design stages, from the needs analysis to design verification, have been executed under the defined processes and activities. Moreover, this work contributes two significant design areas for digitalized ASTS. These are firstly, how to categorize the ASTS if the ASTS has a backed up function of the manual reactor trip, and secondly, how to set the requirements using the given design practices either in overseas ASTS design or similar design. In addition, the methodology for determining the setpoint can be applied to the I and C design and development project which needs to justify the error sources correctly. The systematic approach that has been developed and realized in this work can be utilized in designing new I and C (instrument and control system) as well.

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures

Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

International Nuclear Information System (INIS)

Brooks, B.G.

1985-03-01

This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

International Nuclear Information System (INIS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-01-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK registered platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed

Pyrolysis of rice husk and corn stalk in auger reactor:Part 1. Characterization of char and gas at various temperatures

OpenAIRE

Yu, Yang; Yang, Yang; Cheng, Zhicai; Blanco, Paula H.; Liu, Ronghou; Bridgwater, A.V.; Cai, Junmeng

2016-01-01

In this study, rice husk and corn stalk have been pyrolyzed in an auger pyrolysis reactor at pyrolysis temperatures of 350, 400, 450, 500, 550, and 600 °C in order to investigate the effect of the pyrolysis temperature on the pyrolysis performance of the reactor and physicochemical properties of pyrolysis products (this paper focuses on char and gas). The results have shown that the pyrolysis temperature significantly affects the mass yields and properties of the pyrolysis products. The mass ...

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Partial nitrification using aerobic granules in continuous-flow reactor: rapid startup.

Science.gov (United States)

Wan, Chunli; Sun, Supu; Lee, Duu-Jong; Liu, Xiang; Wang, Li; Yang, Xue; Pan, Xiangliang

2013-08-01

This study applied a novel strategy to rapid startup of partial nitrification in continuous-flow reactor using aerobic granules. Mature aerobic granules were first cultivated in a sequencing batch reactor at high chemical oxygen demand in 16 days. The strains including the Pseudoxanthomonas mexicana strain were enriched in cultivated granules to enhance their structural stability. Then the cultivated granules were incubated in a continuous-flow reactor with influent chemical oxygen deamnad being stepped decreased from 1,500 ± 100 (0-19 days) to 750 ± 50 (20-30 days), and then to 350 ± 50 mg l(-1) (31-50 days); while in the final stage 350 mg l(-1) bicarbonate was also supplied. Using this strategy the ammonia-oxidizing bacterium, Nitrosomonas europaea, was enriched in the incubated granules to achieve partial nitrification efficiency of 85-90% since 36 days and onwards. The partial nitrification granules were successfully harvested after 52 days, a period much shorter than those reported in literature. Copyright © 2013 Elsevier Ltd. All rights reserved.

Graphite for high-temperature reactors

International Nuclear Information System (INIS)

Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

1976-01-01

The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

31 CFR 500.572 - Humanitarian projects authorized.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Humanitarian projects authorized. 500.572 Section 500.572 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued... from private sources, including but not limited to accredited degree-granting institutes of education...

13 CFR 146.500 - Secretary of Defense.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Secretary of Defense. 146.500 Section 146.500 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION NEW RESTRICTIONS ON LOBBYING..., a covered Federal action from the prohibition whenever the Secretary determines, in writing, that...

40 CFR 152.500 - Requirements for devices.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Requirements for devices. 152.500 Section 152.500 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) PESTICIDE PROGRAMS... other than a bacterium, virus, or other microorganism on or in living man or living animals) but not...

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Study on thermodynamic cycle of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Qu Xinhe; Yang Xiaoyong; Wang Jie

2017-01-01

The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

A review of fast reactor progress in Japan, March 1979

Energy Technology Data Exchange (ETDEWEB)

Tomabechi, K

1979-07-01

The fast reactor development project in Japan will be continued in the next fiscal year, from April 1979 through March 1980, at a similar scale of effort both in budget and personnel, to those of the fiscal year of 1978. The total budget for LMFBR development for the next fiscal year is approximately 24 billion Yen, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast reactor development in the PNC is approximately 500, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, approval for power increase from presently approved 50 MWt to 75 MWt with the present core and also to 100 MWt with a modified core in the future was granted by the regulatory authority in September 1978. Two operational cycles at 50 MWt have been completed very recently and preparation for power increase to 75 MWt is being made. With respect to the prototype fast breeder reactor MONJU, progress toward construction is being made and an environmental impact statement of MONJU filed last autumn is being reviewed by the concerned authorities. By the new atomic energy law recently made effective in Japan, the tasks of the former Japan Atomic Energy Commission were split into two and the Atomic Energy Safety Commission was newly established on 4th October 1978 in order to deal with nuclear safety problems in the country. All other problems are treated by the Atomic Energy Commission, as before. Highlights and topics of the fast reactor development activities in the past twelve months are summarized in this paper.

Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

International Nuclear Information System (INIS)

Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

2015-01-01

At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

24 CFR 954.500 - Repayment of investment.

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Repayment of investment. 954.500... DEVELOPMENT INDIAN HOME PROGRAM Program Administration § 954.500 Repayment of investment. (a) HOME funds will be made available pursuant to a HOME Investment Partnership Agreement. The agreement ensures that...

31 CFR 500.325 - National securities exchange.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false National securities exchange. 500.325... Definitions § 500.325 National securities exchange. The term national securities exchange shall mean an exchange registered as a national securities exchange under section 6 of the Securities Exchange Act of...

Final-Independent Confirmatory Survey Report For The Reactor Building, Hot Laboratory, Primary Pump House, And Land Areas At The Plum Brook Reactor Facility, Sandusky, Ohio DCN:2036-SR-01-10

International Nuclear Information System (INIS)

Bailey, Erika N.

2011-01-01

In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities

Differentiation of pulmonary embolism from high altitude pulmonary edema

International Nuclear Information System (INIS)

Khan, D.A.; Hashim, R.; Mirza, T.M.; Matloob-ur-Rehman, M.

2003-01-01

Objective: To differentiate the high altitude pulmonary edema (HAPE) from pulmonary embolism (PE) by clinical probability model of PE, lactate dehydrogenase (LDH), aspartate transaminase (AST) and D-dimer assays at high altitude. Subjects and Methods: Consecutive 40 patients evacuated from height > 3000 meters with symptoms of PE or HAPE were included. Clinical pretest probabilities scores of PE, Minutex D-dimer assay (Biopool international) and cardiac enzymes estimation by IFCC approved methods, were used for diagnosis. Mann-Whitney U test was applied by using SPSS and level of significance was taken at (p 500 ng/ml. Plasma D-dimer of 500 ng/ml was considered as cut-off value; 6(66.7%) patients of PE could be diagnosed and 30 (96.7%) cases of HAPE excluded indicating very good negative predictive value. Serum LDH, AST and CK were raised above the reference ranges in 8 (89%), 7 (78%) and 3 (33%) patients of PE as compared to 11 (35%), 6 (19%) and 9 (29%) of HAPE respectively. Conclusion: Clinical assessment in combination with D-dimer assay, LDH and AST can be used for timely differentiation of PE from HAPE at high altitude where diagnostic imaging procedures are not available. (author)

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

13 CFR 500.100 - Purpose and scope.

Science.gov (United States)

2010-01-01

... Board's authorities and organizational structure, the means and rules by which the Board takes actions... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Purpose and scope. 500.100 Section 500.100 Business Credit and Assistance EMERGENCY OIL AND GAS GUARANTEED LOAN BOARD EMERGENCY OIL AND...

13 CFR 500.213 - Termination of obligations.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Termination of obligations. 500.213 Section 500.213 Business Credit and Assistance EMERGENCY OIL AND GAS GUARANTEED LOAN BOARD... the Application, the Guarantee or the Loan Documents; (5) A Lender fails to make a demand for payment...

49 CFR 179.500-10 - Protective housing.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Protective housing. 179.500-10 Section 179.500-10...-10 Protective housing. (a) Safety devices, and loading and unloading valves on tanks shall be protected from accidental damage by approved metal housing, arranged so it may be readily opened to permit...

TOP500 Supercomputers for November 2004

Energy Technology Data Exchange (ETDEWEB)

Strohmaier, Erich; Meuer, Hans W.; Dongarra, Jack; Simon, Horst D.

2004-11-08

24th Edition of TOP500 List of World's Fastest Supercomputers Released: DOE/IBM BlueGene/L and NASA/SGI's Columbia gain Top Positions MANNHEIM, Germany; KNOXVILLE, Tenn.; BERKELEY, Calif. In what has become a closely watched event in the world of high-performance computing, the 24th edition of the TOP500 list of the worlds fastest supercomputers was released today (November 8, 2004) at the SC2004 Conference in Pittsburgh, Pa.

TOP500 Supercomputers for June 2004

Energy Technology Data Exchange (ETDEWEB)

Strohmaier, Erich; Meuer, Hans W.; Dongarra, Jack; Simon, Horst D.

2004-06-23

23rd Edition of TOP500 List of World's Fastest Supercomputers Released: Japan's Earth Simulator Enters Third Year in Top Position MANNHEIM, Germany; KNOXVILLE, Tenn.;&BERKELEY, Calif. In what has become a closely watched event in the world of high-performance computing, the 23rd edition of the TOP500 list of the world's fastest supercomputers was released today (June 23, 2004) at the International Supercomputer Conference in Heidelberg, Germany.

TOP500 Supercomputers for June 2002

Energy Technology Data Exchange (ETDEWEB)

Strohmaier, Erich; Meuer, Hans W.; Dongarra, Jack; Simon, Horst D.

2002-06-20

19th Edition of TOP500 List of World's Fastest Supercomputers Released MANNHEIM, Germany; KNOXVILLE, Tenn.;&BERKELEY, Calif. In what has become a much-anticipated event in the world of high-performance computing, the 19th edition of the TOP500 list of the worlds fastest supercomputers was released today (June 20, 2002). The recently installed Earth Simulator supercomputer at the Earth Simulator Center in Yokohama, Japan, is as expected the clear new number 1. Its performance of 35.86 Tflop/s (trillions of calculations per second) running the Linpack benchmark is almost five times higher than the performance of the now No.2 IBM ASCI White system at Lawrence Livermore National Laboratory (7.2 Tflop/s). This powerful leap frogging to the top by a system so much faster than the previous top system is unparalleled in the history of the TOP500.

31 CFR 500.407 - Administration of blocked estates of decedents.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Administration of blocked estates of decedents. 500.407 Section 500.407 Money and Finance: Treasury Regulations Relating to Money and Finance... Interpretations § 500.407 Administration of blocked estates of decedents. Section 500.201 prohibits all...

Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

Energy Technology Data Exchange (ETDEWEB)

Kukreja, Mukesh [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)]. E-mail: mrkukreja@yahoo.com

2005-08-01

Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India.

Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

International Nuclear Information System (INIS)

Kukreja, Mukesh

2005-01-01

Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

Low cost, high yield IFE reactors: Revisiting Velikhov's vaporizing blankets

International Nuclear Information System (INIS)

Logan, B.G.

1992-01-01

The performance (efficiency and cost) of IFE reactors using MHD conversion is explored for target blanket shells of various materials vaporized and ionized by high fusion yields (5 to 500 GJ). A magnetized, prestressed reactor chamber concept is modeled together with previously developed models for the Compact Fusion Advanced Rankine II (CFARII) MHD Balance-of-Plant (BoP). Using conservative 1-D neutronics models, high fusion yields (20 to 80 GJ) are found necessary to heat Flibe, lithium, and lead-lithium blankets to MHD plasma temperatures, at initial solid thicknesses sufficient to capture most of the fusion yield. Advanced drivers/targets would need to be developed to achieve a ''Bang per Buck'' figure-of-merit approx-gt 20 to 40 joules yield per driver $ for this scheme to be competitive with these blanket materials. Alternatively, more realistic neutronics models and better materials such as lithium hydride may lower the minimum required yields substantially. The very low CFARII BoP costs (contributing only 3 mills/kWehr to CoE) allows this type of reactor, given sufficient advances that non-driver costs dominate, to ultimately produce electricity at a much lower cost than any current nuclear plant

31 CFR 500.511 - Transactions by certain business enterprises.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Transactions by certain business enterprises. 500.511 Section 500.511 Money and Finance: Treasury Regulations Relating to Money and Finance... Licenses, Authorizations and Statements of Licensing Policy § 500.511 Transactions by certain business...

Considerations regarding design of ion exchange columns for applications in heavy water nuclear reactors- a comprehensive review

International Nuclear Information System (INIS)

Joginder Kumar; Nema, M.K.

2000-01-01

In nuclear reactor applications the principal role of the purification system is to maintain a satisfactory chemistry of moderator and coolant which are different at various stages of reactor operations e.g. during reactor start up, for removal of neutron poison from the moderator, the purification flows are much different compared to steady state operation of the reactor. In order to cater to varying requirements regarding purification load, optimisation in connection with ion exchange column design plays an important role and becomes very challenging in Heavy Water Nuclear Reactors mainly due to the fact that heavy water is very very expensive. In this paper a comprehensive review is made for various designs adopted so far regarding IX column in Indian PHWRs of 220 MWe size for normal operations. Design and operating experience regarding large size IX column used for occasional needs during dilute chemical decontamination of 220 MWe PHWRs is also discussed. The experience regarding development testing of the proposed design of ion exchange column for 500 MWe PHWRs is also discussed

Example of End States of Decommissioning Phases from the Decommissioning of the Multipurpose Research Reactor MZFR, Karlsruhe, Germany

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-06-15

The multipurpose reactor MZFR was a pressurized water reactor, cooled and moderated with heavy water. It was built from 1961 to 1966, and went critical for the first time on 29 September 1965. After 19 years of successful operation, the reactor was shut down on 3 May 1984. The reactor had a thermal output of 200 MW, and an electrical output of 50 MW. In addition to generating electricity, the MZFR had the following functions: - Testing fuel assemblies and various materials for reactor construction; - Gaining experience in the design, erection and operation of heavy water reactor systems; - Training scientific and technical reactor personnel; - Providing heat (first nuclear combined heat and power system (1979-1984)). In 1989, it was decided to dismantle the reactor completely, step by step. The decommissioning concept for the plant, down to a greenfield site, provides for eight distinct decommissioning steps (phases). A separate decommissioning licence was required for each step. The decommissioning work was carried out according to pre-approved work schedules. About 72 000 t of concrete and 7200 t of metal were to be removed. About 1000 t of concrete (500 t biological shield) and 1680 t of metal were to be classified as radioactive waste.

49 CFR 179.500-6 - Heat treatment.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Heat treatment. 179.500-6 Section 179.500-6...-6 Heat treatment. (a) Each necked-down tank shall be uniformly heat treated. Heat treatment shall... treatment of alternate steels shall be approved. All scale shall be removed from outside of tank to an...

31 CFR 500.529 - Powers of attorney.

Science.gov (United States)

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Powers of attorney. 500.529 Section 500.529 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... section does not authorize the creation of any power of attorney in favor of any person outside of the...

49 CFR 179.500-7 - Physical tests.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Physical tests. 179.500-7 Section 179.500-7...-7 Physical tests. (a) Physical tests shall be made on two test specimens 0.505 inch in diameter... tank. These test specimen ring sections or prolongations shall be heat treated, with the necked-down...

Career Development Programs in Fortune 500 Firms.

Science.gov (United States)

Keller, Jack; Piotrowski, Chris

Career development programs (CDPs) are a rather recent area of study in organizational and industrial psychology. The present study investigated the nature and evaluation of CDPs in Fortune 500 firms. Data were obtained by a mailed questionnaire completed by the firms' human resources directors. Of the 500 companies surveyed, only those 50 that…

Effects of irradiation on ferritic alloys and implications for fusion reactor applications

International Nuclear Information System (INIS)

Gelles, D.S.

1986-07-01

This paper reviews the ADIP irradiation effects data base on ferritic (martensitic) alloys to provide reactor teams with an understanding of how such alloys will behave for fusion reactor first wall applications. Irradiation affects dimensional stability, strength and toughness. Dimensional stability is altered by precipitation and void swelling. Swelling as high as 25% may occur in some ferritic alloys at 500 dpa. Irradiation alters strength both during and following irradiation. Irradiation at low temperatures leads to hardening whereas at higher temperatures and high exposures, precipitate coarsening can result in softening. Toughness can also be adversely affected by irradiation. Failure can occur in ferritic in a brittle manner and irradiation induced hardening causes brittle failure at higher temperatures. Even at high test temperatures, toughness is reduced due to reduced failure initiation stresses. 39 refs

Temperature reactivity coefficient of the RA reactor; Temperaturni koeficijenat reaktivnosti reaktora RA

Energy Technology Data Exchange (ETDEWEB)

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

1961-12-15

Temperature reactivity coefficient of the RA reactor was determined as follows. Stabilization of moderator temperature and graphite reflector was achieved in the reactor operated at power levels of 20, 100, 500, 1000, 3000 and 5000 kW. Temperature change of the moderator was achieved by changing the water flow rate in the secondary cooling system. The fuel temperature was changed simultaneously. During the measurement at each power level the temperature change was between 30 - 50 deg C. Changing the position of the automated regulator is registered during moderator temperature change, and these changes were used for determining the total reactivity change by using the calibration curves for the automated regulator. In the measured temperature range the the reactivity change was linear and it was possible to determine the total temperature coefficient.

TOP500 Supercomputers for June 2005

Energy Technology Data Exchange (ETDEWEB)

Strohmaier, Erich; Meuer, Hans W.; Dongarra, Jack; Simon, Horst D.

2005-06-22

25th Edition of TOP500 List of World's Fastest Supercomputers Released: DOE/L LNL BlueGene/L and IBM gain Top Positions MANNHEIM, Germany; KNOXVILLE, Tenn.; BERKELEY, Calif. In what has become a closely watched event in the world of high-performance computing, the 25th edition of the TOP500 list of the world's fastest supercomputers was released today (June 22, 2005) at the 20th International Supercomputing Conference (ISC2005) in Heidelberg Germany.

29 CFR 500.81 - Payment of wages when due.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Payment of wages when due. 500.81 Section 500.81 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR REGULATIONS MIGRANT AND SEASONAL AGRICULTURAL WORKER PROTECTION Worker Protections Wages and Payroll Standards § 500.81 Payment of...

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

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

CERN Document Server

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

2016-11-08

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

Basic study on high temperature gas cooled reactor technology for hydrogen production

International Nuclear Information System (INIS)

Chang, Jong Hwa; Lee, W. J.; Lee, H. M.

2003-01-01

The annual production of hydrogen in the world is about 500 billion m 3 . Currently hydrogen is consumed mainly in chemical industries. However hydrogen has huge potential to be consumed in transportation sector in coming decades. Assuming that 10% of fossil energy in transportation sector is substituted by hydrogen in 2020, the hydrogen in the sector will exceed current hydrogen consumption by more than 2.5 times. Currently hydrogen is mainly produced by steam reforming of natural gas. Steam reforming process is chiefest way to produce hydrogen for mass production. In the future, hydrogen has to be produced in a way to minimize CO2 emission during its production process as well as to satisfy economic competition. One of the alternatives to produce hydrogen under such criteria is using heat source of high-temperature gas-cooled reactor. The high-temperature gas-cooled reactor represents one type of the next generation of nuclear reactors for safe and reliable operation as well as for efficient and economic generation of energy

Ferritin levels, inflammatory biomarkers, and mortality in peripheral arterial disease: a substudy of the Iron (Fe) and Atherosclerosis Study (FeAST) Trial.

Science.gov (United States)

Depalma, Ralph G; Hayes, Virginia W; Chow, Bruce K; Shamayeva, Galina; May, Patricia E; Zacharski, Leo R

2010-06-01

This study delineated correlations between ferritin, inflammatory biomarkers, and mortality in a cohort of 100 cancer-free patients with peripheral arterial disease (PAD) participating in the Veterans Affairs (VA) Cooperative Study #410, the Iron (Fe) and Atherosclerosis Study (FeAST). FeAST, a prospective, randomized, single-blind clinical trial, tested the hypothesis that reduction of iron stores using phlebotomy would influence clinical outcomes in 1227 PAD patients randomized to iron reduction or control groups. The effects of statin administration were also examined in the Sierra Nevada Health Care (SNHC) cohort by measuring serum ferritin levels at entry and during the 6-year study period. No difference was documented between treatment groups in all-cause mortality and secondary outcomes of death plus nonfatal myocardial infarction and stroke. Iron reduction in the main study caused a significant age-related improvement in cardiovascular disease outcomes, new cancer diagnoses, and cancer-specific death. Tumor necrosis factor (TNF)-alpha, TNF-alpha receptors 1 and 2, interleukin (IL)-2, IL-6, IL-10, and high-sensitivity C reactive protein (hs-CRP) were measured at entry and at 6-month intervals for 6 years. Average levels of ferritin and lipids at entry and at the end of the study were compared. The clinical course and ferritin levels of 23 participants who died during the study were reviewed. At entry, mean age of entry was 67 +/- 9 years for the SNHCS cohort, comparable to FeAST and clinical and laboratory parameters were equivalent in substudy participants randomized to iron reduction (n = 51) or control (n = 49). At baseline, 53 participants on statins had slightly lower mean entry-level ferritin values (114.06 ng/mL; 95% confidence interval [CI] 93.43-134.69) vs the 47 off statins (127.62 ng/mL; 95% CI, 103.21-152.02). Longitudinal analysis of follow-up data, after adjusting for the phlebotomy treatment effect, showed that statin use was associated with

42 CFR 456.500 - Purpose.

Science.gov (United States)

2010-10-01

... ASSISTANCE PROGRAMS UTILIZATION CONTROL Utilization Review Plans: FFP, Waivers, and Variances for Hospitals and Mental Hospitals § 456.500 Purpose. For hospitals and mental hospitals, this subpart— (a...

Stress analysis of secondary ramp and secondary tilting mechanism of inclined fuel transfer machine for 500 MWe PFBR

International Nuclear Information System (INIS)

Prabhakaran, K.M.; Vaze, K.K.; Ghosh, A.K.; Rai, Somesh; Sundarani, A.R.; Patel, R.J.; Agrawal, R.G.

2004-10-01

Inclined Fuel Transfer Machine (IFTM) is one of the important machine of the fuel handling system of 500 MWe Prototype Fast Breeder Reactor (PFBR). It is used to transfer core sub-assemblies (CSA) from reactor vessel to fuel building and vice-versa. Secondary ramp and Secondary tilting mechanism (SR/STM) is a part of IFTM which acts as a passage to transfer CSA. This mechanism and components were designed by the Refuelling Technology Division of BARC as per the ASME design code as class 2 component. Being critical in nature and complicated in geometry it was required to check the design of these components by detailed finite element analysis. The loading considered in the present study was static, thermal and seismic conditions. This was done using FEM software COSMOS/M. The Stresses were categorised as per the requirement of the ASME code for various levels of loading (Level A, B and C). Based on the analysis performed, it was concluded that the SR/STM qualifies the requirement of ASME code Section-III NC (Class-2 components). This report gives the details of the studies performed. (author)

25 CFR 11.500 - Law applicable to civil actions.

Science.gov (United States)

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false Law applicable to civil actions. 11.500 Section 11.500 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR LAW AND ORDER COURTS OF INDIAN OFFENSES AND LAW AND ORDER CODE Civil Actions § 11.500 Law applicable to civil actions. (a) In all civil cases, the...

Study on enhancement of heat transfer of reactor vessel auxiliary cooling system of fast breeder reactor

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Kinoshita, Izumi; Ueda, Nobuyuki; Furuya, Masahiro

1996-01-01

A reactor vessel auxiliary cooling system (RVACS), which is one of the decay heat removal systems of the fast breeder reactor (FBR), has passive safety as well as high reliability. However, the heat removal capability is relatively small, because its heat exchange is dependent on the natural convection of the air. The objectives of this report are to propose a heat transfer medium to enhance the heat transfer and to confirm the heat transfer performance of this system by experimental and analytical studies. From these studies, the following main results were obtained. (1) A porous plate with 5 mm thickness, 5 mm pore diameter, 92% porosity, was found to have the highest enhancement of heat transfer. (2) The heat transfer enhancement was demonstrated by large scale heat transfer experiments. Also, the heat transfer correlations, which can be used in the plant transient analyses, were derived from the experimental results. (3) Analysing the transient conditions of conventional pool-type FBR by means of the system analysis code, the applicable range of this system was assumed from the capability of the RVACS with porous plates. As a result, this type of RVACS was found to be applicable to conventional pool-type FBRs with capacity of about 500 MWe or less. (author)

Experimental investigation of pyrolysis of rice straw using bench-scale auger, batch and fluidized bed reactors

International Nuclear Information System (INIS)

Nam, Hyungseok; Capareda, Sergio C.; Ashwath, Nanjappa; Kongkasawan, Jinjuta

2015-01-01

Energy conversion efficiencies of three pyrolysis reactors (bench-scale auger, batch, and fluidized bed) were investigated using rice straw as the feedstock at a temperature of 500 °C. The highest bio-oil yield of 43% was obtained from the fluidized bed reactor, while the maximum bio-char yield of 48% was obtained from the batch reactor. Similar bio-oil yields were obtained from the auger and batch type reactors. The GCMS and FTIR were used to evaluate the liquid products from all reactors. The best quality bio-oil and bio-char from the batch reactor was determined to have a heating value of 31 MJ/kg and 19 MJ/kg, respectively. The highest alkali mineral was found in the bio-char produced from the auger reactor. The energy conversion efficiencies of the three reactors indicated that the majority of the energy (50–64%) was in the bio-char products from the auger and batch reactors, while the bio-oil from the fluidized bed reactor contained the highest energy (47%). A Sankey diagram has been produced to show the flows of product energy from each pyrolysis process. The result will help determine which conversion process would be optimal for producing specific products of bio-char, bio-oil, and gas depending on the needs. - Highlights: • Pyrolysis products from auger, batch, and fluidized bed reactor were examined. • O/C ratios of bio-oils stayed in specific ranges depending on the process reactors. • The largest quantity of bio-oil from fluidized, while the best quality from batch. • The highest alkali concentration of 37 g/kg included in the auger based bio-char. • Sankey diagram was used to understand the energy distribution from reactors.

Status of fast breeder reactor development in India

International Nuclear Information System (INIS)

Bhoje, S.B.

1996-01-01

The energy scenario and economic conditions in India are presented. India needs considerable energy for its rapid industrialisation with the liberal economic policy. Nuclear energy with FBR is the only large scale energy resource other than coal, available in the country. The present economic constraints have delayed the construction of new NPPs. The performance of operating reactors has improved considerably during the year. Operating experience of FBTR has been detailed particularly the reactivity incident and its investigations. Updated design of 500 MWe PFBR is presented. Various R and D works in support of FBR in the engineering, metallurgy, chemistry, reprocessing, safety etc. are detailed. (author)

Status of fast breeder reactor development in India

Energy Technology Data Exchange (ETDEWEB)

Bhoje, S B [Reactor Group, IGCAR, Kalpakkam (India)

1996-07-01

The energy scenario and economic conditions in India are presented. India needs considerable energy for its rapid industrialisation with the liberal economic policy. Nuclear energy with FBR is the only large scale energy resource other than coal, available in the country. The present economic constraints have delayed the construction of new NPPs. The performance of operating reactors has improved considerably during the year. Operating experience of FBTR has been detailed particularly the reactivity incident and its investigations. Updated design of 500 MWe PFBR is presented. Various R and D works in support of FBR in the engineering, metallurgy, chemistry, reprocessing, safety etc. are detailed. (author)

Creep behavior and evolution of microstructure of modified Grade 91 welded joint after short term exposure at 500 deg C

International Nuclear Information System (INIS)

Vivier, F.

2009-03-01

With the increase in worldwide energy demand, the nuclear industry is a way of producing electricity on a large scale and to answer to this need. For the design of a new generation of fission nuclear reactors and among six chosen fission reactor systems, France develops in particularly the Very High Temperature Reactor (VHTR) concept. This implies the use of materials that are more and more resistant to high temperature for long-term exposure. AREVA focuses on materials already used in fossil-fuel power plant, so that the mechanical behaviour of Grade 91 (Fe 9 Cr 1 MoNbV) has to be investigated. This ferritic-martensitic steel is considered to be a potential candidate for welded components. Such structures are combined with welded joints, which have to be studied. Three industrial partners (AREVA, CEA, EDF) have launched a study with the Centre des Materiaux in order to investigate the creep of welded joint of Grade 91. The aim of this work is to complete the available database about the mechanical behaviour of Grade 91, base metal and welded joint, during creep tests performed at 500 C up to 4500 h exposure. Thermal aging tests, tensile tests, and creep tests were performed at 450 C and 500 C using both base metal and cross-weld samples. Several geometries of cross-weld creep specimens were tested. The microstructure has not remarkably changed after tests concerning both nature and size of precipitates, and the characteristic size of the matrix sub-structure. The creep damage is not developed in the ruptured specimens after creep tests. Only little damage by cavity nucleation and growth was found in the creep specimens. Creep fracture at 500 C takes places by viscoplastic flow, contrary to tests performed at 625 C where the creep-induced damage governs the creep rupture at least for long-term lifetime. From creep curves of base metal and cross-weld specimens, a phenomenological model is proposed. The flow rule is a Norton power law with a stress exponent of 19 in

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Demonstration of the 500 kW stoker burning system; 500 kW stokeripolttojaerjestelmaen demonstrointi

Energy Technology Data Exchange (ETDEWEB)

Oravainen, H. [VTT Energy, Jyvaeskylae (Finland); Kantalainen, K. [Hoegfors Laempoe Oy, Saarijaervi (Finland)

1995-12-31

The objective of the project is to demonstrate the operation of a 500 kW stoker-burning system in practice. The project is continuation of the previous projects of the Bioenergy research programme, 303 - Development of automatic heating system for wood chips and sod peat, carried out by VTT Energy, and Y301 - Development of heating boiler for wood chips and sod peat, carried out by Hoegfors Laempoe Oy. A 500 kW (nominal power) stoker-burner was constructed on the basis of the results of these projects. The burner was mounted on ETNA 500 bioenergy boiler. Screw-feeders, constructed by Maamiehen Saehkoe Oy, were used as fuel feeding system. Maamiehen Saehkoe Oy delivered also the automation system for the combustion equipment. Combustion air pre-heater was mounted on the boiler for promotion of the combustion of moist fuel. Testing of the equipment was carried out at the laboratory of VTT Energy in Jyvaeskylae in October-November 1994. In December 1994 the equipment was transported to Jalasjaervi, to heating station of the farmer Juha Jyrae. The actual heat generation started in the beginning of February 1995, when the greenhouses started to require heating. Sod peat has been used as the fuel. VTT Energy has carried out the efficiency and emission measurements in the heating station with sod peat in March 1995, and with reed canary grass in autumn 1995. The energy generation and fuel consumption have been followed all the time

Demonstration of the 500 kW stoker burning system; 500 kW stokeripolttojaerjestelmaen demonstrointi

Energy Technology Data Exchange (ETDEWEB)

Oravainen, H [VTT Energy, Jyvaeskylae (Finland); Kantalainen, K [Hoegfors Laempoe Oy, Saarijaervi (Finland)

1996-12-31

The objective of the project is to demonstrate the operation of a 500 kW stoker-burning system in practice. The project is continuation of the previous projects of the Bioenergy research programme, 303 - Development of automatic heating system for wood chips and sod peat, carried out by VTT Energy, and Y301 - Development of heating boiler for wood chips and sod peat, carried out by Hoegfors Laempoe Oy. A 500 kW (nominal power) stoker-burner was constructed on the basis of the results of these projects. The burner was mounted on ETNA 500 bioenergy boiler. Screw-feeders, constructed by Maamiehen Saehkoe Oy, were used as fuel feeding system. Maamiehen Saehkoe Oy delivered also the automation system for the combustion equipment. Combustion air pre-heater was mounted on the boiler for promotion of the combustion of moist fuel. Testing of the equipment was carried out at the laboratory of VTT Energy in Jyvaeskylae in October-November 1994. In December 1994 the equipment was transported to Jalasjaervi, to heating station of the farmer Juha Jyrae. The actual heat generation started in the beginning of February 1995, when the greenhouses started to require heating. Sod peat has been used as the fuel. VTT Energy has carried out the efficiency and emission measurements in the heating station with sod peat in March 1995, and with reed canary grass in autumn 1995. The energy generation and fuel consumption have been followed all the time

Noteworthy: Fortune 500: Texas ties California for national lead

OpenAIRE

Michael Nicholson

2010-01-01

According to the 2010 Fortune 500, released in April, Texas hosts the headquarters of 57 of the nation's 500 largest companies, ranked by gross revenues. Texas secured its place as a Fortune 500 leader through its position as focal point of the domestic energy industry, its relatively strong economic growth over the past decade, and its relatively low tax rates and living costs.

Fuel Management at the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Pham, V.L.; Nguyen, N.D.; Luong, B.V.; Le, V.V.; Huynh, T.N.; Nguyen, K.C. [Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City (Viet Nam)

2011-07-01

The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The spent fuel storage was newly designed and installed in the place of the old thermalizing column for biological irradiation. The core was loaded by Russian WWR-M2 fuel assemblies (FAs) with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 highly enriched uranium (HEU) FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. After the shuffling the working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam Atomic Energy Institute the mixed core configurations of irradiated HEU and new low enriched uranium (LEU) FAs has been created on 12 September, 2007 and on 20 July, 2009. After reloading in 2009, the 14 HEU FAs with highest burnup were removed from the core and put in the interim storage in reactor pool. The works on full core conversion for the DNRR are being realized in cooperation with the organizations, DOE and IAEA. Contract for Nuclear fuel manufacture and supply of 66 LEU FAs for DNRR

LOCA Analysis of KAIST-Micro Modular Reactor with Modified GAMMA+ code

Energy Technology Data Exchange (ETDEWEB)

Oh, Bong Seong; Ahn, Yoon Han; Kim, Seong Gu; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

The supercritical carbon dioxide (S-CO{sub 2}) power cycle is being seriously investigated around the world due to its simple layout, quite high efficiency around 500 .deg. C turbine inlet temperature, etc. By combining these two ideas, the KAIST research team developed a S-CO{sub 2} cooled SMR, called KAIST-Micro Modular reactor (MMR), which is targeting transportability and electricity supply for remote region. Therefore, requirements of MMR design are factory fabrication of the total system including power conversion system to be transported and air cooling to be independent from the site selection. Until now, steady performances and sizes of components were evaluated. Thus, in this paper a transient performance of the MMR are simulated with special focus on the loss of coolant accident (LOCA) at cold leg pipe. The MMR is a newly suggested innovative small modular reactor concept by the KAIST research team. Since the MMR is cooled by supercritical CO{sub 2}, general safety codes for conventional reactors have limitations. Thus, GAMMA+ code for the transient analysis of a gas-cooled reactor was selected and modified for the S-CO{sub 2} power system. After the modification of GAMMA+ code, LOCA is simulated, which is considered as one of the most limiting accidents in terms of safety of nuclear power plant.

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

International Nuclear Information System (INIS)

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

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

Pyrolysis of Rubber in a Screw Reactor

Science.gov (United States)

Lozhechnik, A. V.; Savchin, V. V.

2016-11-01

On the basis of an analysis of thermal methods described in the literature and from the results of experimental investigations of steam conversion, the authors have developed and created a facility for thermal processing of rubber waste. Rubber crumb was used as the raw material; the temperature in the reactor was 500°C; nitrogen, steam, and a mixture of light hydrocarbons (noncondensable part of pyrolysis products) represented the working medium. The pyrolysis yielded 36-38% of a solid fraction, 54-56% of a liquid hydrocarbon fraction, and 6-9% of noncondensable gases. Changes in the composition of the gas mixture have been determined at different stages of processing. Gas chromatography of pyrolysis gases has shown that the basic gases produced by pyrolysis are H2 and hydrocarbons C2H4, C3H6, C3H8, C4H8, C2H6, C3H6O2, and C4H10, and a small amount of H2S, CO, and CO2. Noncondensable gases will be used as a fuel to heat the reactor and to implement the process.

Fast pyrolysis of corn stovers with ceramic ball heat carriers in a novel dual concentric rotary cylinder reactor.

Science.gov (United States)

Fu, Peng; Bai, Xueyuan; Li, Zhihe; Yi, Weiming; Li, Yongjun; Zhang, Yuchun

2018-05-09

Fast pyrolysis of corn stovers with ceramic ball heat carriers in a dual concentric rotary cylinder reactor was studied to explore the product yields and characteristics in response to temperature. The reactor was confirmed to successfully scale up to a 25 kg/h pilot plant, with its performance being excellent. The highest bio-oil yield of 48.3 wt% at 500 °C was attained with the char and gas yields being 26.8 and 24.9 wt%. Phenols content was reduced from 22.3% to 18.9% when elevating temperature from 450 until 600 °C, with guaiacols and alkyl phenols being the predominant compounds, while ketones accounted for 15.8-23.0% and their content showed a continuous increase, with hydroxyacetone being the paramount ketonic one. Acetic acid was the dominant acidic compound with its peak content of 9.4% at 500 °C. The char characteristics in response to temperatures were determined for subsequent processing and high value-added utilization. Copyright © 2018 Elsevier Ltd. All rights reserved.

Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

International Nuclear Information System (INIS)

Sanatkumar, A.; Jit, I.; Muralidhar, G.

1996-01-01

India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs

Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

Energy Technology Data Exchange (ETDEWEB)

Sanatkumar, A; Jit, I; Muralidhar, G [Nuclear Power Corporation of India Ltd., Mumbai (India)

1997-12-31

India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs.

Production of Radioisotopes and Radiopharmaceuticals at the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Duong Van Dong; Pham Ngoc Dien; Bui Van Cuong; Mai Phuoc Tho; Nguyen Thi Thu; Vo Thi Cam Hoa

2014-01-01

After reconstruction, the Dalat Nuclear Research Reactor (DNRR) was inaugurated on March 20th, 1984 with the nominal power of 500 kW. Since then the production of radioisotopes and labelled compounds for medical use was started. Up to now, DNRR is still the unique one in Vietnam. The reactor has been operated safely and effectively with the total of about 37,800 hrs (approximately 1,300 hours per year). More than 90% of its operation time and over 80% of its irradiation capacity have been exploited for research and production of radioisotopes. This paper gives an outline of the radioisotope production programme using the DNRR. The production laboratory and facilities including the nuclear reactor with its irradiation positions and characteristics, hot cells, production lines and equipment for the production of Kits for labelling with 99m Tc and for quality control, as well as the production rate are mentioned. The methods used for production of 131 I, 99m Tc, 51 Cr, 32 P, etc. and the procedures for preparation of radiopharmaceuticals are described briefly. Status of utilization of domestic radioisotopes and radiopharmaceuticals in Vietnam is also reported. (author)

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)