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Sample records for bw standard reactor

  1. MPOWER (B&W Generation mPower Inc., USA) [Passive Safety Systems in Advanced Small Modular Reactors

    International Nuclear Information System (INIS)

    The B&W mPower™ reactor module is an integral PWR designed by B&W to generate an output of 180 MW(e). The inherent safety features of the reactor design include a low core linear heat rate which reduces fuel and cladding temperatures during accidents, a large reactor coolant system volume which allows more time for safety system responses in the event of an accident, and small penetrations at high elevations, increasing the amount of coolant available to mitigate a small break LOCA. The emergency core cooling system is connected with the reactor coolant inventory purification system and removes heat from the reactor core after anticipated transients in a passive manner, while also passively reducing containment pressure and temperature. The plant is designed without taking credit for safety related emergency diesel generators, and a design objective is no core uncovering during design basis accidents

  2. Standards for reactor accident cases

    International Nuclear Information System (INIS)

    The Committee on Standards for reactor accident cases in the Netherlands published its recommendations to the Minister of Health. Maximum permissible quantities of radiation and radionuclide intake are presented for adults and children as well as pregnant women. Exposure limit standards for the whole body as well as specific organs and other parts are given. Also considered is the contamination of cattle and cows' milk. The standards are compared with those of the ICRP and the English Medical Research Council

  3. Verdivurdering av BW Offshore

    OpenAIRE

    Kalleberg, Carl-Fredrik; Storebø, Mats

    2015-01-01

    Målet med oppgaven var å gjennomføre en verdivurdering av BW Offshore. Problemstillingen var å finne egenkapitalverdien for en privat investor i BW Offshore og deretter komme med en handelsanbefaling på BWO aksjen. For å gjennomføre verdivurderingen valgte vi en fundamentalanalyse ved bruk av den diskonterte kontantstrømmodellen. Vi supplerte verdivurderingen med en komparativ- og sensitivitetsanalyse. BW Offshore opererer innenfor exploration & production (E&P) i offshorebr...

  4. New revisions of reactor physics standards

    International Nuclear Information System (INIS)

    This paper presents an overview of two of the most basic reactor physics standards, the newly revised “Steady State Neutronics Methods for Power Reactor Analysis” and its companion standard, “Nuclear Data for Reactor Design”. These two popular state-of-the-art standards provide important guidance for developing the necessary input data needed to calculate reactor lattice and core parameters such as reaction rates spatial distributions, reactivity and flux distributions in power reactors for all currently used reactor types, from fast to thermal reactors. The standards provide guidance for the selection of cross section data and libraries, the development of nuclear data sets suitable for specific applications, energy group structures and group collapsing. Key elements in the complex sequence of power reactor calculations are outlined. The effects of simplifications and approximations made in the treatment of the neutronic and geometric models and the biases and uncertainties resulting from such simplifications and assumptions are discussed. In the newly-revised standard on neutronics methods, clear distinction is made between the two important tools for assessing the reliability of the results of the calculations, verification and validation. To provide an auditable path in the verification and validation processes, the standard requires detailed documentation including methods used, selection of calculation models and experimental data and results of higher order calculations. (author)

  5. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  6. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

  7. Reactor Section standard analytical methods. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Sowden, D.

    1954-07-01

    the Standard Analytical Methods manual was prepared for the purpose of consolidating and standardizing all current analytical methods and procedures used in the Reactor Section for routine chemical analyses. All procedures are established in accordance with accepted practice and the general analytical methods specified by the Engineering Department. These procedures are specifically adapted to the requirements of the water treatment process and related operations. The methods included in this manual are organized alphabetically within the following five sections which correspond to the various phases of the analytical control program in which these analyses are to be used: water analyses, essential material analyses, cotton plug analyses boiler water analyses, and miscellaneous control analyses.

  8. IAEA role in nuclear reactor safety standardization

    International Nuclear Information System (INIS)

    In 1981 the electricity generation by nuclear power plants all over the world reached 8% of total production. It can be expected that at the turn of century up to 25% of electric power will be provided by means of nuclear fuel burning. In connection with NPP total number growth, their attraction to large population centres, widening of the atomic energy application areas, the importance of nuclearreactor safety problems can only increase. The safety measures have usually the structure of sequential barriers: for accident preventing, for protection from accidents, for accident localization. NPP safety is a complex problem having scientific, engineering, juridical, social and political aspects. Since these problems have an international importance, IAEA should actively work on their solving. Practically all the topics of nuclear power development and nuclear reactor s;fety lie within the activity area of the Department of nuclear power and safety, its sections: of nuclear safety, nuclear power, nuclear fuel cycle. In 1974 a decision was made in IAEA about initiation of work on development of an international nuclear safety standards system (NUSS Programme). These activities are divided into five major branches: a government organization for nuclear safety regulations; site selection for NPP; NPP desing; operation, start of operation and decommissioning; quality provision for NPP. The report presents a list of documents, comprising the NUSS Programme. The complection of all the works within the scope of the Programme is planned for 1985. After 1985 the start of development of fast neutron reactor and fuel cycle enterprise safety standards is planned

  9. Brief overview of American Nuclear Society's research reactor standards

    International Nuclear Information System (INIS)

    The American Nuclear Society (ANS) established the research reactor standards group in 1968. The standards group, known as ANS-15, was established for the purpose of developing, preparing, and maintaining standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training

  10. American Nuclear Society standards for TRIGA reactors and their use

    International Nuclear Information System (INIS)

    The American Nuclear Society established a committee (ANS-15) with the expressed charter to develop standards for research reactors. These standards were to cover all aspects of research reactor operations, maintenance and administration. Standards have been written in every area of research reactor operations that the research reactor community has deemed important. One of the uppermost goals of the Standards Committee work is to produce standards that provide guidance and help to the research reactor community in a timely manner. To make the standards meaningful requires a great deal of cooperation between all segments of the reactor community. The research reactors - whether they are private, university or government owned - have a mission to perform. At the same time, the regulatory agencies also have a mission to perform, and with a spirit of mutual respect and cooperation, both can accomplish their goals. In the last five years this spirit has been present, and a number of very good standards have resulted. These standards should be a part of every research reactor library. In particular ANS-15.16 and ANS-15.1 have been endorsed by the regulatory agencies and are being used to evaluate submittals

  11. Integrated lid unit for a nuclear reactor of standard construction

    International Nuclear Information System (INIS)

    This is an integrated lid unit for a nuclear reactor of standard construction, where many components and sub-groups of the upper reactor structure are collected into one unit, which is lifted in one lifting operation from the reactor containment vessel. The integrated lid unit includes, in particular, the pressure vessel lid, a cooling jacket, the control rod drive mechanisms, a catch plate, a lifting device, a winch and a cable connection plate. (orig.)

  12. International standardization of nuclear reactor designs - the way forward

    International Nuclear Information System (INIS)

    The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

  13. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  14. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  15. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-05-22

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

  16. Reference design for the standard mirror hybrid reactor

    International Nuclear Information System (INIS)

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (239Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket

  17. Outlines of revised regulation standards for experimental research reactors

    International Nuclear Information System (INIS)

    In response to the accident of TEPCO Fukushima Daiichi Nuclear Power Station, the government took actions through the revision of regulatory standards as well as the complete separation of regulation administrative department from promotion administrative department. The Nuclear and Industrial Safety Agency of the Ministry of Economy, Trade and Industry, which has been in charge of the regulations of commercial reactors, and the Office of Nuclear Regulations of the Ministry of Education, Culture, Sports, Science and Technology, which has been in charge of the regulations of reactors for experiment and research, were separated from both ministries, and integrated into the Nuclear Regulation Authority, which was newly established as the affiliated agency of the Ministry of the Environment. As for the revision of regulations and standards, the Nuclear Safety Commission was dismantled, and regulation enacting authority was given to the new Nuclear Regulation Authority, and the regulations that stipulated new regulatory standards were enacted. This paper outlines the contents of regulations related mainly to the reactors for experiment and research, and explains the following: (1) retroactive application of the new regulatory standards to existing reactor facilities, (2) examinations at the Nuclear Regulatory Agency, (3) procedures to confirm the compliance to the new standards, (4) seismic design classification, and (5) importance classification of safety function. (A.O.)

  18. Fusion reactor design studies: standard accounts for cost estimates

    International Nuclear Information System (INIS)

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

  19. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  20. Among farm variation in heifer BW gains.

    Science.gov (United States)

    Bond, G B; von Keyserlingk, M A G; Chapinal, N; Pajor, E A; Weary, D M

    2015-11-01

    BW of replacement heifers is rarely measured on commercial farms, making it difficult to evaluate the success of management practices related to calf growth. Our aims were to describe variability among commercial farms in Holstein heifer BW, determine how BW differences varied with management and propose a method of estimating calf growth based upon single measurement. Heart girth circumference was used to estimate BW of 576 heifers 48 to 70 weeks of age on 33 different farms (on average 11 ± 6 heifers/farm) in British Columbia, Canada. Regression analysis showed a linear relationship of BW with age (BW (kg)=116+5 × age (weeks)). Residuals from this regression were averaged across heifers within each farm to identify farms where heifers were heavier or lighter than would be predicted on the basis of their age; farm average residuals ranged from -54 to 72 kg. Farms with heifers showing the highest residual BW also had the highest rates of gain for pre-weaned calves. These results indicate that farms able to rear faster growing calves before weaning were also rearing faster growing heifers at breeding, and suggest that management of milk-fed calves is a particularly important component of replacement heifer management. PMID:26477529

  1. Guidance of clearance related standards in reactor facilities

    International Nuclear Information System (INIS)

    The reactor regulation law was amended in May 2005 to provide for the clearance system. The regulatory body confirmed radioactivity concentration of material and allowed the material be classified as 'material cleared from regulatory control'. The procedure of confirmation consisted of two steps; 1) methodology and 2) results, of measurement and assessment. Clearance related standards were issued as ordinance of Ministry and NISA information notice, which specified radioactive nuclides, radioactivity concentration, evaluation unit, how to decide radioactivity concentration, radiation measurement equipment and control of object material. This report was guidance of clearance related standards in reactor facilities with explanatory notes. By applying the clearance system, material of insignificant radiation level, which was part of concrete or metals generated from decommissioning of reactor facilities, could be released from regulatory control of radioactive waste as clearance material, and recycle of such material could result in reduction of radioactive waste amount. (T. Tanaka)

  2. Neutron standard cross sections in reactor physics - Need and status

    International Nuclear Information System (INIS)

    The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community

  3. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  4. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    International Nuclear Information System (INIS)

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience

  5. Non-Power Reactor Operator Licensing Examiner Standards

    International Nuclear Information System (INIS)

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR Part 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, this standard will be revised periodically to accommodate comments and reflect new information or experience

  6. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  7. On the path to ordering standardized advanced light water reactors

    International Nuclear Information System (INIS)

    The international Advanced Light Water Reactor (ALWR) program is specifying, designing, and certifying the next generation of nuclear power plants. Begun in the mid-1980's, the program is on track to permit ordering and construction of families of standardized plants at the start of the twenty-first century. ALWRs will be constructed only if they are economically competitive with alternative forms of electricity generation and are recognized as acceptable and favorable by the public, prospective owners, and investors. This paper first gives an overview of the major building blocks ensuring safe, reliable, and economic designs and the status of those designs. Next it lays out the path the industry has charted toward adopting the ALWR option and indicates the status of three key steps -- design certification, utility requirements, and first-of-a-kind engineering. Lastly, the paper focuses on one of the most important building blocks for ensuring economic viability -- life-cycle standardization. Among the topics are the definition and scope of standardization; its advantages and disadvantages; design team standardization plans that describe the desired or optimum degree of standardization and the processes used to achieve it; and the need for an agreement among all plant owners and operators for implementing and sustaining standardization in families of ALWRs. 10 refs., 5 figs

  8. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  9. BW Trained HMM based Aerial Image Segmentation

    Directory of Open Access Journals (Sweden)

    R Rajasree

    2011-03-01

    Full Text Available Image segmentation is an essential preprocessing tread in a complicated and composite image dealing out algorithm. In segmenting arial image the expenditure of misclassification could depend on the factual group of pupils. In this paper, aggravated by modern advances in contraption erudition conjecture, I introduce a modus operandi to make light of the misclassification expenditure with class-dependent expenditure. The procedure assumes the hidden Markov model (HMM which has been popularly used for image segmentation in recent years. We represent all feasible HMM based segmenters (or classifiers as a set of points in the beneficiary operating characteristic (ROC space. optimizing HMM parameters is still an important and challenging work in automatic image segmentation research area. Usually the Baum-Welch (B-W Algorithm is used to calculate the HMM model parameters. However, the B-W algorithm uses an initial random guess of the parameters, therefore after convergence the output tends to be close to this initial value of the algorithm, which is not necessarily the global optimum of the model parameters. In this project, a Adaptive Baum-Welch (GA-BW is proposed.

  10. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  11. Utilization of Research Reactors in Standard Reference Material Certification

    Energy Technology Data Exchange (ETDEWEB)

    Capannesi, G.; Rosada, A. [UTFISST-CATNUC, ENEA, R.C.-Casaccia, via Anguillarese 301, 00060 Rome (Italy); Avino, P. [DIPIA, INAIL (ex-ISPESL), via Urbana 167, 00184 Rome (Italy)

    2011-07-01

    The certification issue of Standard Reference Materials is one of the most complex analytical problems and runs over different research fields. International organization, e.g. NIST, BCR etc., organize continuously systematic intercomparison campaigns among worldwide laboratories using different analytical techniques. Samples are irradiated in nuclear research reactors and analyzed by Instrumental Neutron Activation Analysis, a technique strongly involved in this field for its significant analytical properties. This paper shows a study on Zircaloy-4. The importance of accurate measurements of minor constituents, i.d. Cr, Fe, Hf and Sn, regards its characteristics of corrosion resistance and mechanical properties. The samples were irradiated in the rotating rack of the TRIGA Mark II reactor of the R.C.-Casaccia (ENEA). The gamma spectrometry measurements were performed after 30 and 90 days of decay by means of HPGe detector. The results obtained by interlaboratory intercomparison can highlight an excellent precision for Cr, Hf and Sn, and a good precision for Fe. The reliability of the technique is confirmed by Hf determination, since the INAA is one of the few analytical techniques measuring and delivering accurate and homogeneous data. (author)

  12. Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1...

  13. General principles of nuclear reactor instrumentation (International Electrotechnical Commission Standard Publication 60231:1967)

    International Nuclear Information System (INIS)

    This Recommendation given guidance on the provision of reactor instrumentation and recommends standards of good practice. The main body of the Recommendation is of general application and aspects applicable only to particular types of reactors are included in Appendices. Items of instrumentation are included only where they have a direct bearing on the over-all safety and effective control of the reactor

  14. Radiation streaming analysis in the korean standard nuclear power plant reactor cavity

    International Nuclear Information System (INIS)

    Radiation shield plugs are provided in the reactor cavity of the KSNP (Korean Standard Nuclear Power Plant) to assure an acceptable low level of radiation streaming up through the reactor cavity. An analysis of radiation streaming in the reactor cavity was performed to evaluate the effectiveness of the reactor cavity shield plugs in meeting their design goals. The MCNP code was used for this analysis with DORT leakage flux at the reactor vessel outer surface. Based on the results of this analysis, the upper and lower shield plugs are important design features to reduce dose rates at the reactor vessel flange level and at the operating floor

  15. The metamorphosis of SN1998bw

    CERN Document Server

    Patat, F; Danziger, J; Mazzali, P A; Sollerman, J; Augusteijn, T; Brewer, J; Doublier, V; González, J F; Hainaut, O R; Lidman, C E; Leibundgut, B; Nomoto, K; Nakamura, T; Spyromilio, J; Rizzi, L; Turatto, M; Walsh, J; Galama, T J; Van Paradijs, J; Kouveliotou, C; Vreeswijk, P M; Frontera, F; Masetti, N; Palazzi, E; Pian, E; Patat, Ferdinando; Cappellaro, Enrico; Danziger, John; Mazzali, Paolo A.; Sollerman, Jesper; Augusteijn, Thomas; Brewer, James; Doublier, Vanessa; Gonzalez, Jean Francois; Hainaut, Olivier; Lidman, Chris; Leibundgut, Bruno; Nomoto, Ken'ichi; Nakamura, Takayoshi; Spyromilio, Jason; Rizzi, Luca; Turatto, Massimo; Walsh, Jeremy; Galama, Titus J.; Paradijs, Jan van; Kouveliotou, Chryssa; Vreeswijk, Paul M.; Frontera, Filippo; Masetti, Nicola; Palazzi, Eliana; Pian, Elena

    2001-01-01

    We present and discuss the photometric and spectroscopic evolution of the peculiar SN1998bw, associated with GRB980425, through an analysis of optical and near IR data collected at ESO-La Silla. The spectroscopic data, spanning the period from day -9 to day +376 (relative to B maximum), have shown that this SN was unprecedented, although somewhat similar to SN1997ef. Maximum expansion velocities as high as 3x10^4 km/s to some extent mask its resemblance to other Type Ic SNe. At intermediate phases, between photospheric and fully nebular, the expansion velocities (~10^4 km/s) remained exceptionally high compared to those of other recorded core-collapse SNe at a similar phase. The mild linear polarization detected at early epochs suggests the presence of asymmetry in the emitting material. The degree of asymmetry, however, cannot be decoded from these measurements alone. The HeI 1.083 mu and 2.058 mu lines are identified and He is suggested to lie in an outer region of the envelope. The temporal behavior of the...

  16. Standard technical specifications for Westinghouse pressurized water reactors (revision issued Fall 1981). Technical report

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

  17. Standard interface files and procedures for reactor physics codes. Version IV

    International Nuclear Information System (INIS)

    Standards, procedures, and recommendations of the Committee on Computer Code Coordination for promoting the exchange of reactor physics codes are updated to Version IV status. Standards and procedures covering general programming, program structure, standard interface files, and file management and handling subroutines are included

  18. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  19. Experience of the standardization of the vibratory condition pipe line when working the reactor on powers

    International Nuclear Information System (INIS)

    Analysis of the experience of the standardization of the vibratory condition pipe line and considered approaches of the motivation of the normative requirements is organized in article to vibratory load on pipe lines when working the reactor on powers

  20. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

  1. IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

  2. EST Table: BW999031 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999031 SES2029 11/12/09 n.h 10/09/28 52 %/242 aa ref|NP_001136335.1| egalitarian ... [Nasonia vitr ... gi|91084227|ref|XP_969046.1| PREDICTED: similar to egalitarian ... CG4051-PA [Tribolium castaneum] BW999031 L8 ...

  3. Choosing a standard reactor: International competition and domestic politics in Chinese nuclear policy

    International Nuclear Information System (INIS)

    China has ambitious plans to expand its nuclear power capacity. One of the policy goals that high-level policymakers have desired is to base the nuclear program on a standardized reactor design. However, this has not materialized so far. By examining its nuclear reactor choices for individual projects, we argue that China’s policymaking process has been greatly influenced by international competition and domestic politics. Multiple international nuclear vendors are intent upon maintaining their respective niches in the expanding Chinese reactor market, and they have used various forms of economic and political pressure to achieve their objectives. On the other hand, China’s policymaking process is fragmented and the shifting power balances among powerful domestic actors do not allow a fixed path to be followed. Further, because of the high costs and potential profits involved, nuclear reactor choices in China have been driven not just by technical considerations but also by foreign and trade policy objectives. All of these make it unlikely that China will standardize the reactor type it constructs in the near future. -- Highlights: ► China’s nuclear power policymaking has been fragmented and without central control. ► Multiple domestic actors have pursued independent agendas. ► International nuclear vendors have intensely competed for Chinese reactor contracts. ► Economic, political and foreign policy goals have driven reactor contract decisions. ► China is unlikely to construct only a standardized reactor design.

  4. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  5. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    Energy Technology Data Exchange (ETDEWEB)

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  6. Operator licensing examination standards for power reactors. Interim revision 8

    International Nuclear Information System (INIS)

    These examination standards are intended to assist NRC examiners and facility licensees to better understand the processes associated with initial and requalification examinations. The standards also ensure the equitable and consistent administration of examinations for all applicants. These standards are for guidance purposes and are not a substitute for the operator licensing regulations (i.e., 10 CFR Part 55), and they are subject to revision or other changes in internal operator licensing policy. This interim revision permits facility licensees to prepare their initial operator licensing examinations on a voluntary basis pending an amendment to 10 CFR Part 55 that will require facility participation. The NRC intends to solicit comments on this revision during the rulemaking process and to issue a final Revision 8 in conjunction with the final rule

  7. EST Table: BW999064 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999064 SES2141 11/12/09 GO hit GO:0005524(ATP binding)|GO:0006754(ATP biosynthetic process)|GO ... ase activity, coupled to transmembrane movement of ions , phosphorylative mechanism)|GO:0016020(membrane) 1 ...

  8. Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  9. Standard technical specifications for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on General Electric plants currently being reviewed for an Operating License. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. This revision of the GE-STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  10. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  11. Compiled reports on the applicability of selected codes and standards to advanced reactors

    International Nuclear Information System (INIS)

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC

  12. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  13. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Final report. Volume 1

    International Nuclear Information System (INIS)

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included: (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other ongoing programs. This information included: (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  14. Nuclear reactors' construction costs: The role of lead-time, standardization and technological progress

    International Nuclear Information System (INIS)

    This paper provides the first comparative analysis of nuclear reactor construction costs in France and the United States. Studying the cost of nuclear power has often been a challenge, owing to the lack of reliable data sources and heterogeneity between countries, as well as the long time horizon which requires controlling for input prices and structural changes. We build a simultaneous system of equations for overnight costs and construction time (lead-time) to control for endogeneity, using expected demand variation as an instrument. We argue that benefits from nuclear reactor program standardization can arise through short term coordination gains, when the diversity of nuclear reactors' technologies under construction is low, or through long term benefits from learning spillovers from past reactor construction experience, if those spillovers are limited to similar reactors. We find that overnight construction costs benefit directly from learning spillovers but that these spillovers are only significant for nuclear models built by the same Architect-Engineer (A- E). In addition, we show that the standardization of nuclear reactors under construction has an indirect and positive effect on construction costs through a reduction in lead-time, the latter being one of the main drivers of construction costs. Conversely, we also explore the possibility of learning by searching and find that, contrary to other energy technologies, innovation leads to construction costs increases. (authors)

  15. Benchmark calculations by the thermal reactor standard nuclear design code system SRAC

    International Nuclear Information System (INIS)

    This report summarizes the present status of the thermal reactor standard nuclear design code system SRAC developed by the nuclear design working group of the JAERI thermal reactor standard code committee which was started on July 1978. Descriptions are given at first on the brief introduction and the process of development of the code system SRAC, and then, the several benchmark tests performed to evaluate the performance of the code system. The results show the good predictions of the experimental keff values of the critical facilities; TCA for LWR, JMTRC for JAERI MTR, DCA for the Japanese Advanced Thermal Reactor and SHE for VHTR. A trial to the IAEA benchmark calculations on the Reduction of uranium Enrichment of Research and Test Reactors yields satisfactory agreements with the results of ANL. Another test to evaluate the fast group constants was also attempted by tracing the fast reactor benchmark problems which have been used to evaluate nuclear data file in the FBR reactor physics field. (author)

  16. High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.; Swearingen, Gary L.; Pugh, Ray

    2011-10-31

    Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC

  17. Standard technical specifications for Babcock and Wilcox pressurized water reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. The format of the STS addresses the categories required by 10 CFR 50 and consists of six sections covering the areas of: Definitions, Safety Limits and Limiting Safety System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls

  18. Standard technical specifications for Babcock and Wilcox pressurized water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  19. Standard technical specifications for combustion engineering pressurized water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Combustion Engineering plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  20. Standard technical specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Westinghouse plants currently being reviewed for an Operating License. Accordingly, the document contains specifications applicable to plants with (1) either 3 or 4 loops and (2) with and without loop stop valves. In addition, four separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, Ice Condenser, Sub-Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of the STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  1. Nuclear reactors' construction costs: The role of lead-time, standardization and technological progress

    International Nuclear Information System (INIS)

    This paper provides an econometric analysis of nuclear reactor construction costs in France and the United States based on overnight costs data. We build a simultaneous system of equations for overnight costs and construction time (lead-time) to control for endogeneity, using change in expected electricity demand as instrument. We argue that the construction of nuclear reactors can benefit from standardization gains through two channels. First, short term coordination benefits can arise when the diversity of nuclear reactors' designs under construction is low. Second, long term benefits can occur due to learning spillovers from past constructions of similar reactors. We find that construction costs benefit directly from learning spillovers but that these spillovers are only significant for nuclear models built by the same Architect–Engineer. In addition, we show that the standardization of nuclear reactors under construction has an indirect and positive effect on construction costs through a reduction in lead-time, the latter being one of the main drivers of construction costs. Conversely, we also explore the possibility of learning by searching and find that, contrary to other energy technologies, innovation leads to construction costs increases. -- Highlights: •This paper analyses the determinants of nuclear reactors construction costs and lead-time. •We study short term (coordination gains) and long term (learning by doing) benefits of standardization in France and the US. •Results show that standardization of nuclear programs is a key factor for reducing construction costs. •We also suggest that technological progress has contributed to construction costs escalation

  2. An inspection standard of fuel for the high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author).

  3. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  4. Standard review plan for the review and evaluation of emergency plans for research and test reactors

    International Nuclear Information System (INIS)

    This document provides a Standard Review Plan to assure that complete and uniform reviews are made of research and test reactor radiological emergency plans. The report is organized under ten planning standards which correspond to the guidance criteria in American National Standard ANSI/ANS 15.16 - 1982 as endorsed by Revision 1 to Regulatory Guide 2.6. The applicability of the items under each planning standard is indicated by subdivisions of the steady-state thermal power levels at which the reactors are licensed to operate. Standard emergency classes and example action levels for research and test reactors which should initiate these classes are given in an Appendix. The content of the emergency plan is as follows: the emergency plan addresses the necessary provisions for coping with radiological emergencies. Activation of the emergency plan is in response to the emergency action levels. In addition to addressing those severe emergencies that will fall within one of the standard emergency classes, the plan also discusses the necessary provisions to deal with radiological emergencies of lesser severity that can occur within the operations boundary. The emergency plan allows for emergency personnel to deviate from actions described in the plan for unusual or unanticipated conditions

  5. Mechanical Proofs about BW Multi-Party Contract Signing Protocol

    Institute of Scientific and Technical Information of China (English)

    ZHANG Ningrong; ZHANG Xingyuan; WANG Yuanyuan

    2006-01-01

    We report on the verification of a multi-party contract signing protocol described by Baum-Waidner and Waidner (BW). Based on Paulson' inductive approach, we give the protocol model that includes infinitely many signatories and contract texts signing simultaneously. We consider composite attacks of the dishonest signatory and the external intruder, formalize cryptographic primitives and protocol arithmetic including attack model, show formal description of key distribution, and prove signature key secrecy theorems and fairness property theorems of the BW protocol using the interactive theorem prover Isabelle/HOL.

  6. Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    Return to Contents page 1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in derived neutron dosimetry results. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  7. EST Table: BW999584 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999584 SES4246 10/09/28 62 %/102 aa sp|P38404.1|GNAO_LOCMI RecName: Full=Guanine nucleotide-bi ... 8 58 %/102 aa C26C6.2#CE05311#WBGene00001648#locus:goa -1#guanine nucleotide-binding protein#status:Confir ...

  8. EST Table: BW998950 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998950 SES1744 10/09/28 67 %/161 aa ref|XP_969609.1| PREDICTED: similar to microfibr...ene00009671#locus:mfap-1#MICROFIBRILLAR- ASSOCIATED PROTEIN 1 (ASSOCIATED MICROFIBR...aa gi|91077510|ref|XP_969609.1| PREDICTED: similar to microfibrillar-associated protein 1 [Tribolium castaneum] AU005329 L8 ...

  9. EST Table: BW997643 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997643 ES1616 10/09/28 45 %/133 aa ref|XP_969699.1| PREDICTED: similar to fuzzy C...10/09/10 45 %/133 aa gi|91081315|ref|XP_969699.1| PREDICTED: similar to fuzzy CG13396-PA [Tribolium castaneum] CK514734 L8 ...

  10. EST Table: BW998335 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998335 ES4403 10/09/28 69 %/146 aa ref|NP_726360.3| egalitarian ... [Drosophila melanogaster] gb|A ... AF47054.4| egalitarian ... [Drosophila melanogaster] 10/08/29 69 %/146 aa FBp ... gi|91084227|ref|XP_969046.1| PREDICTED: similar to egalitarian ... CG4051-PA [Tribolium castaneum] FS766799 L8 ...

  11. EST Table: BW999494 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999494 SES3749 10/09/28 65 %/182 aa ref|XP_972095.2| PREDICTED: similar to rap55 [Tribolium ca ... 0 %/116 aa Y18D10A.17#CE21413#WBGene00012484#locus:car - 1#status:Confirmed#UniProt:Q9XW17#protein_id:CAA2 ...

  12. EST Table: BW998051 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998051 ES3190 10/09/28 62 %/146 aa ref|XP_002091851.1| GE13875 [Drosophila yakuba] gb|EDW91563 ... 48 %/146 aa F58B3.5b#CE44702#WBGene00003415#locus:mars - 1#status:Partially_confirmed#UniProt:D3YT55#prote ...

  13. EST Table: BW999096 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999096 SES2266 10/09/28 85 %/193 aa ref|NP_001040402.1| preimplantation protein [...Bombyx mori] gb|ABF51322.1| preimplantation protein [Bombyx mori] 10/08/29 73 %/195 aa FBpp0252287|DwilGK231

  14. The IAEA Safety Standards and Small and Medium Reactor Design Features

    International Nuclear Information System (INIS)

    The renewed interest in nuclear energy worldwide prompted the development of small and medium size reactors (SMR). These reactors are more suitable for smaller grids as well as they promise increased safety, possibly better economy, and they offer flexibility in applications (including non-power), siting and fuel cycle options. The IAEA safety standards, especially the Fundamental Safety Principles (IAEA Safety Standards series No. SF-1) provide excellent principles for design of nuclear power systems, these principles however, as such, are very generic. The IAEA safety standard Safety of Nuclear Power Plants: Design (SSR-2/1) provides specific requirements for the design of nuclear power plants, yet these requirements were developed based on best practices related to the design of current generation of reactors. On other hand, the SMRs currently under design incorporate features that are innovative but with which there is no operational experience, therefore the current standards and design requirement might not be directly applicable to the new SMR designs. This paper examines four most advanced in their development designs (NuScale, mPower, SMART, CAREM) with respect to design safety requirements of the IAEA. These designs were selected because all of them are pressurized water reactors and among SMRs they are closest relatives to the current designs currently Being offered to the market. Therefore, the existing safety standards should be potentially applicable. Since these SMR designs are under development and most of the detailed information is not available this paper utilizes only information that is public domain. This limits the depth of the evaluation and mainly principal technical requirements and some general design requirements are addressed. Focus is on issues such as defence in depth, design principles (single failure criterion, fail safe design, etc.), design extension conditions, external hazards. (author)

  15. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

  16. Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

    CERN Document Server

    Li, Yu-Feng

    2014-01-01

    We discuss reactor antineutrino oscillations with non-standard interactions (NSIs) at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline) dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO) as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters.

  17. Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

    International Nuclear Information System (INIS)

    We discuss reactor antineutrino oscillations with non-standard interactions (NSIs) at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline) dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO) as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters

  18. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to June 1978

  19. Three-dimensional flow field analysis of the standard fuel assembly for China advanced research reactor

    International Nuclear Information System (INIS)

    Numerical simulation of the flow field of the standard fuel assembly in China Advanced Research Reactor is carried out by using computational fluid dynamics software CFX4.4 and CFX5.5. The flow distribution and pressure difference of different coolant channels in the standard fuel assembly at rated operating condition are reached. Based on the computational pressure drop results of different flow rate, the resistance characteristic curve is given and compared with experimental results. The two results are in good agreement. (authors)

  20. Standardization of the SLOWPOKE-2 reactor in Jamaica for routine NAA

    International Nuclear Information System (INIS)

    The International Centre for Environmental and Nuclear Sciences (ICENS) has been involved in conducting multipurpose geochemical surveys, the results of which were published in 'A Geochemical Atlas of Jamaican Soils'. The primary analytical tool for these studies was neutron activation analysis (NAA) using the SLOWPOKE-2 reactor at the Centre. The neutron flux of the SLOWPOKE-2 reactor is extremely stable, thus allowing a semi-absolute method for quantitative NAA. This has several advantages, but requires preparation and measurement of the single- or multi-element standards for each gamma-spectroscopy system (GSS). The NAA laboratory at ICENS operates three GSSs. The primary ('master') GSS was standardized using single element standards for over 50 elements, naturally occurring in most geological and biological materials. The standardization of the secondary GSS's was achieved by transferring of the elemental sensitivities of the master GSS using an instrumentation-free standardization approach. Implementation of this methodology and its utilization in the routine analytical work is described. (author)

  1. The cloud storage service bwSync&Share at KIT

    CERN Document Server

    CERN. Geneva

    2014-01-01

    The Karlsruhe Institute of Technology introduced the bwSync&Share collaboration service in January 2014. The service is an on-premise alternative to existing public cloud storage solutions for students and scientists in the German state of Baden-Württemberg, which allows the synchronization and sharing of documents between multiple devices and users. The service is based on the commercial software PowerFolder and is deployed on a virtual environment to support high reliability and scalability for potential 450,000 users. The integration of the state-wide federated identity management system (bwIDM) and a centralized helpdesk portal allows the service to be used by all academic institutions in the state of Baden-Württemberg. Since starting, approximately 15 organizations and 8,000 users joined the service. The talk gives an overview of related challenges, technical and organizational requirements, current architecture and future development plans.

  2. SN1998bw/GRB980425 and Radio Supernovae

    CERN Document Server

    Weiler, K W; Montes, M J; Weiler, Kurt W.; Panagia, Nino; Montes, Marcos J.

    2001-01-01

    The unusual supernova SN1998bw, which is thought to be related to the gamma-ray burster GRB980425, is a possible link between the two classes of objects. Analyzing the extensive radio emission data avaliable for SN1998bw, we are able to describe its time evolution within the well established framework available for the analysis of radio emission from supernovae. This then allows description of a number of physical properties of the object. The radio emission can best be explained as interaction of a mildly relativistic (Gamma about 1.6) shock with a dense pre-explosion stellar wind established circumstellar medium (CSM) which is highly structured both azimuthally, in clumps or filaments, and radially, with two observed density enhancements separated by about 3e17 cm. With assumptions as to pre-explosion stellar wind conditions, it is possible to estimate that the progenitor to SN 1998bw had a mass loss rate of about 3.5e-5 solar masses per yr with at least two approximately 30% increases in mass-loss rate; th...

  3. Selective methods for the maintainability and standardization of the engineering of a research reactor

    International Nuclear Information System (INIS)

    the same function in each specialty. These diversities bring about conflicts and confusion between the maintenance and operation crew, besides modifying dangerously the fail rate and thus the overall reliability of the reactor. The maintainability is the capacity of being maintained an equipment/system has, serving as a design parameter. A system must be designed in a way in which it is maintained without a great investment of time and with low costs, minimum environmental impact and the least resources possible. Standardization is the action of normalizing the engineering of all systems/equipments of the reactor from its design, in all the disciplines, (mechanical, electrical, electronic, chemical, etc.) taking into consideration the facility of its maintenance and conserving or increasing the reliability of the system. The intention of this Program of Maintainability and Standardization in Research Reactors is based on procedures and calculations to improve the reliability of the equipments/systems according to pre-established criterion. (author)

  4. Corrosion experiment on non standard austenitic steel A1, in reactor coolant water

    International Nuclear Information System (INIS)

    Experimental corrosion studies on non standard austenitic SS, A1, have been carried out. The samples were immersed in reactor coolant water medium with pH variation of 5.95, 6.0, 6.1, and 6.31. The experiments were carried out using a type of M-273 EG&G potentiostate/galvanometer test instrument. The post-corrosion samples' microstructure were analyzed with the aid of EDS (energy dispersive spectroscopy) equipped SEM instrument to detect the presence of any viable corrosion products. For further verification x-ray diffraction method was also used to detect any possible emerging corrosion products type on the samples' surfaces. Experimental results confirm that non standard austenitic SS immersed in reactor coolant water corrosion medium with a variation of concentration experience very little or almost no corrosion, and that according to the so-called Fontana's criteria these test-materials turn out to have an excellent resistance toward reactor coolant water corrosion medium. This is also evidenced by the very low corrosion rate value measured in this study. EDS study and X-ray diffraction results indicate that the possible ensuing corrosion by products are chrome oxides and iron oxides. (author)

  5. Knowledge management - A key issue for EnBW

    International Nuclear Information System (INIS)

    Full text: The motivation for knowledge management can be summarised with the words of EnBW CEO Prof. Claassen, 2002 'knowledge manager of the year' in Germany: 'Against the backdrop of the ever-increasing complexity of strategic planning and activities on the operational front, knowledge management is a key factor in the long-term success of our business.' Professional knowledge management motivates and supports employees, helping to create networks in which they can lay the foundations for the future success of the company. It must be emphasised that knowledge management is the responsibility of management, and EnBW has established a suitable framework consisting of different action levels and goals: 1) Normative level (corporate culture): creating a knowledge-aware and knowledge-friendly corporate culture 2) Strategic level (human resources): systematic gearing of internal intangible potentials towards future requirements. 3) Operational level (information/communication): making the required knowledge available in the necessary scope and quality, in the right place and at the right time. If knowledge management can generally be seen as basic requirement for successful companies, then it is even more important for nuclear operators. Today, nuclear energy is an important generating technology in Europe, for Germany and for EnBW with major future potential, but a technology that must be employed with great caution and attaching top priority to safe operation. For nuclear operators, the rule is always 'safety first'. But knowledge management implemented and used in the right way can also enhance both safety and competitive operation of the plants at the same time. In this connection, successful knowledge management plays a key role due to the complex interplay of many different disciplines within a demanding legal and regulatory framework, the paramount importance of collective past experience and the high demands on the expertise of the employees operating nuclear

  6. Longitudinal analysis of residual feed intake and BW in mink using random regression with heterogeneous residual variance.

    Science.gov (United States)

    Shirali, M; Nielsen, V H; Møller, S H; Jensen, J

    2015-10-01

    The aim of this study was to determine the genetic background of longitudinal residual feed intake (RFI) and BW gain in farmed mink using random regression methods considering heterogeneous residual variances. The individual BW was measured every 3 weeks from 63 to 210 days of age for 2139 male+female pairs of juvenile mink during the growing-furring period. Cumulative feed intake was calculated six times with 3-week intervals based on daily feed consumption between weighing's from 105 to 210 days of age. Genetic parameters for RFI and BW gain in males and females were obtained using univariate random regression with Legendre polynomials containing an animal genetic effect and permanent environmental effect of litter along with heterogeneous residual variances. Heritability estimates for RFI increased with age from 0.18 (0.03, posterior standard deviation (PSD)) at 105 days of age to 0.49 (0.03, PSD) and 0.46 (0.03, PSD) at 210 days of age in male and female mink, respectively. The heritability estimates for BW gain increased with age and had moderate to high range for males (0.33 (0.02, PSD) to 0.84 (0.02, PSD)) and females (0.35 (0.03, PSD) to 0.85 (0.02, PSD)). RFI estimates during the growing period (105 to 126 days of age) showed high positive genetic correlations with the pelting RFI (210 days of age) in male (0.86 to 0.97) and female (0.92 to 0.98). However, phenotypic correlations were lower from 0.47 to 0.76 in males and 0.61 to 0.75 in females. Furthermore, BW records in the growing period (63 to 126 days of age) had moderate (male: 0.39, female: 0.53) to high (male: 0.87, female: 0.94) genetic correlations with pelting BW (210 days of age). The result of current study showed that RFI and BW in mink are highly heritable, especially at the late furring period, suggesting potential for large genetic gains for these traits. The genetic correlations suggested that substantial genetic gain can be obtained by only considering the RFI estimate and BW at pelting

  7. Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures. 1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is th...

  8. Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

    International Nuclear Information System (INIS)

    The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

  9. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

    Science.gov (United States)

    Arzu Alpan, F.; Kulesza, Joel A.

    2016-02-01

    This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.

  10. Implementation of k0-standardization method of the INAA at ETRR-2 research reactor

    International Nuclear Information System (INIS)

    The k0-method of INAA standardization has been implemented using the irradiation facilities of the fast pneumatic rabbit and some selected manually loaded irradiation positions, which designated for short and long irradiation, respectively, at Egypt second research reactor. The neutron flux parameters (f and α) in each site have been determined using Zr-Au sets as neutron flux monitors. The reference materials coal NIST 1632c and IAEA-Soil 7 were analyzed for data validation and good agreement between the experimental values and the certified values was obtained. (author)

  11. Standard technical specifications for Babcock and Wilcox pressurized water reactors. Revision 4. Technical report

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors (BandW-STS) is a generic document prepared by the U.S. NRC for use in the licensing process. The BandW-STS provide applicants with model specifications to be used in formulation plant-specific technical specifications required by 10 CFR Part 50, Section 50.36, which set forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection to the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  12. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  13. Utilization of Industrial Borax Wastes (BW) for Portland Cement Production

    OpenAIRE

    ELBEYLİ, İffet YAKAR

    2004-01-01

    Industrial borax wastes (BWs) are formed as solid waste during the production of borax from tincal [Na2B4O5(OH)4.8H2O] in Bandırma, Turkey. These wastes cause different environmental problems and lead to economic losses because of high boron oxide (B2O3) content. The primary aim of this study is the removal of B2O3 from BWs and the second aim is the usage of BWs with low boron content in cement as an additive material. For this purpose, the BW was treated with water for removal of b...

  14. EST Table: BW997255 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997255 ES0256 10/09/28 95 %/115 aa ref|NP_001037584.1| ferritin [Bombyx mori] gb|ABF51339.1| ferrit...in isoform 1 [Bombyx mori] gb|ABG76018.1| iron storage protein [Bombyx mori] gb|ABG76019.1| iron stor...|91077444|ref|XP_966312.1| PREDICTED: similar to putative ferritin 2 isoform 1 [Tribolium castaneum] FS916237 L8 ... ...age protein [Bombyx mandarina] 10/08/29 40 %/113 aa FBpp0225182|DvirGJ10765-PA 10/08/28 n.h 10/09

  15. Study on Kο-standardization method of neutron activation analysis in HANARO research reactor

    International Nuclear Information System (INIS)

    The neutron activation analysis based on kο-standardization method (kο-NAA) has been known as one of the remarkable developments of NAA in recent with advantages of experimental simplicity, high accuracy, excellent flexibility with respect to irradiation and counting conditions, and suitability for computerization. This work has carried out to consider the application of kο-NAA on NAA 1 of HANARO research reactor such as the calibration of gamma-ray spectrometry and irradiation facility. α and f-values related to the reactor neutron spectrum parameters as main factor of irradiation quality are measured and some SRMs are analyzed to compare the accuracy of the measured results. The analytical results with the deviation mostly less than ±15% and all Z-score lower 1.5 in comparison between the experiment value and that value certified of some SRMs has shown a promising capability in application of kο-NAA on the HANARO research reactor

  16. Application of NAA standardization methods using a low power research reactor

    International Nuclear Information System (INIS)

    Two widely used neutron activation analysis (NAA) standardization methods (relative and k0) have been validated at the Ghana Research Reactor-1 (GHARR-1) Centre using environmental and biological standard reference materials (SRMs). The samples were IAEA Soil-7 as an environmental sample, and NIST Orchard Leaves 1571 as a biological sample. The qualitative and quantitative analyses were done using a high resolution Canberra N-type high purity germanium (HPGe) detector. The accuracy and precision were evaluated for the elements analysed. The concentrations of most of the elements were found to be within 10% of the certified values. Precision was calculated from six replicate measurements and was found to be within 15%. (author)

  17. Spectral Evolution of the Peculiar Ic Supernova 1998bw

    CERN Document Server

    Stathakis, R A; Jones, D H; Bessell, M S; Galama, T J; Germany, L M; Hartley, M; James, D M; Kouveliotou, C; Lewis, I J; Parker, Q A; Russell, K S; Sadler, E M; Tinney, C G; Van Paradijs, J; Vreeswijk, P M; Germany, Lisa M.

    2000-01-01

    SN 1998bw holds the record for the most energetic Type Ic explosion, one of the brightest radio supenovae and probably the first supernova associated with a gamma-ray burst. In this paper we present spectral observations of SN 1998bw observed in a cooperative monitoring campaign using the AAT, UKST and the SSO 2.3-m telescope. We investigate the evolution of the spectrum between 7 and 94 days after V_band maximum in comparison to well-studied examples of Type Ic SNe in order to quantify the unusual properties of this supernova event. Though the early spectra differ greatly from the observations of classical Ic SNe, we find that the evolution from the photospheric to the nebular phases is slow but otherwise typical. The spectra differ predominantly in the extensive line blending and blanketing which has been attributed to the high velocity of the ejecta. We find that by day 19, the absorption line minima blueshifts are 10% - 50% higher than other SNe and on day 94 emission lines are 45% broader, as expected if...

  18. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC PROSITE PS00884 OSTEOPONTIN NA ? IPR019841 unintegrated Biol...ogical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  19. Standard review plan for the review and evaluation of emergency plans for research and test reactors. Technical report

    International Nuclear Information System (INIS)

    This document provides a Standard Review Plan for the guidance of the NRC staff to assure that complete and uniform reviews are made of research and test reactor emergency plans. The report is organized under ten planning standards which correspond to the guidance criteria in Draft II of ANSI/ANS 15.16 as endorsed by Revision 1 to Regulatory Guide 2.6. The applicability of the items under each planning standard is indicated by subdivisions of the steady state thermal power levels at which the reactors are licensed to operate. Standard emergency classes and example action levels for research and test reactors which should initiate these classes are given in an Appendix

  20. EST Table: BW998803 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998803 SES1220 10/09/28 50 %/231 aa ref|XP_975194.1| PREDICTED: similar to something about silencing...l|GB19080-PA 10/09/10 50 %/231 aa gi|91081571|ref|XP_975194.1| PREDICTED: similar to something about silencing protein 10 [Tribolium castaneum] FS937283 L8 ... ... protein 10 [Tribolium castaneum] gb|EFA01562.1| hypothetical protein TcasGA2_TC007123 [T...ribolium castaneum] 10/08/29 33 %/236 aa FBpp0231271|DvirGJ16854-PA 10/08/28 n.h 10/09/10 32 %/280 aa AGAP002960-PA Protein...|2R:30130647:30132122:-1|gene:AGAP002960 10/09/10 45 %/223 aa gnl|Ame

  1. Guidelines for estimating present and forecasting future population distributions surrounding power reactor sites. (Draft of a standard)

    International Nuclear Information System (INIS)

    This report presents demographic procedures for estimating and forecasting population totals and distributions within a 50-mile radius of a proposed power plant site. Procedures such as those described are needed to comply with the license application procedure for a proposed power reactor site. The report includes a draft for a prospective American National Standard to estimate present and forecast future population distributions surrounding proposed reactor sites

  2. Catalogue and classification of technical safety standards, rules and regulations for nuclear power reactors and nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    The present report is an up-dated version of the report 'Catalogue and Classification of Technical Safety Rules for Light-water Reactors and Reprocessing Plants' edited under code No EUR 5362e, August 1975. Like the first version of the report, it constitutes a catalogue and classification of standards, rules and regulations on land-based nuclear power reactors and fuel cycle facilities. The reasons for the classification system used are given and discussed

  3. Performance testing of irradiation facility rabbit system pneumatic reactor RSG-GAS using standard reference material

    International Nuclear Information System (INIS)

    The irradiation facility function test of rabbit system pneumatic (RS-5) has been done using standard reference material SRM 1633 Coal Fly ash through the sending station. Long irradiation of about 4-5 seconds. The results of qualitative analysis showed that the dominant elements listed in the certificate can be detected are Al, Ca, Mg, Si, Na, Ti, V, Mn. But only an element of Mn and Na which has a relative refractive values below 10%. And the other elements have a value relative refractive index 25% - 60%. The significant difference of value was not influenced by the position of irradiation in the reactor facility but due to the influence of the time difference between the sample and the standard count, and the half-life nuclide itself. Overall it can be said that the performance of the irradiation facility pneumatic rabbit system is good, but needs to be tested again by using different standard reference materials, in order to obtain the test results of analysis that can be trusted. (author)

  4. Standardization of advanced light water reactors and progress on achieving utility requirements

    International Nuclear Information System (INIS)

    This paper reports that for a number of years, the U.S. utilities had led an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors in the United States. Since 1985, this utility initiative has been effected through a major technical program managed by the Electric Power Research Institute (EPRI); the U.S. Advanced Light Water Reactor (ALWR) Program. In addition to the U.S. utility leadership and sponsorship, the ALWR Program also has the participation and sponsorship of a number of international utility companies and close cooperation with the U.S. Department of Energy (DOE). The NPOC Strategic Plan for Building New Nuclear Plants creates a framework within which new standardized nuclear plants may be built. The Strategic Plan is an expression of the nuclear energy industry's serious intent to create the necessary conditions for new plant construction and operation. The industry has assembled a comprehensive, integrated list of actions that must be taken before new plants will be built and assigns responsibility for managing the various issues and sets time-tables and milestones against which we must measure progress

  5. Reliability study: digital reactor protection system of Korean standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kang, H. G.; Jang, S. C.; Eom, H. S.; Jeong, H. S

    2003-02-01

    Digital safety-critical systems which are now installed in Korean Standard Nuclear Power Plants (KSNPP) would be quantitatively evaluated in order to prove the safety. In this study, we quantify the safety of the digital reactor protection system in KSNPPs using PSA technology. This study also includes the detailed investigation of the target system operation. The Fault Tree (FT) models were constructed for 15 reactor trip parameters. For digital parts, because the operation data for the same type PWR was unavailable, we used the data provided by vendors. On the other hand, for the conventional analog/mechanical parts, we used experience data presented in KAERI/TR-2164/2002.The result of quantification shows that the system unavailability varies from 4.36E-5 to 8.96E-4 according to the trip parameter. Main contributor to the difference from the conventional analysis would be the difference in human failure probability estimation. Generally, the system unavailability depends on several important factors: Human failure probability, software failure probability, watchdog timer coverage, and common cause failure estimation.

  6. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  7. Self absorption and geometric correction factors for reactor off-gas samples relative to NBS standards

    International Nuclear Information System (INIS)

    Although they can be counted in identical bottles using identical counting systems, real gas samples differ from the NBS solution standards (e.g., mock reactor off-gas) in two respects--geometry and self absorption. Because both detector and source are real and finite, the simple ''narrow beam'' linear attenuation coefficient approximations currently used in the industry are quite inadequate for correction. Accordingly, the well-tested, complete-analog program, BIM 130, was used to compute the fraction of photons, and the photon energy spectra, reaching typical detectors used in the industry. Using this method, it was possible to correct the given NBS standard activity to its effective activity relative to a gas sample in an identical bottle. Factors were much closer to unity than predictions based on ''narrow beam'' linear attenuation coefficient approximations. At 80 keV, for example, such approximations gave 0.76, whereas the factor proved to be 1.03 for a 3'' x 3'' NaI(Tl) crystal and a 3 cm distance. Results are presented for various gamma energies of interest from 80 keV to 1,830 keV, and for the commonly used industrial distances of 3, 10, and 30 cm from the bottom of the sample bottle to the top of the detector container. Complete spectra for photons entering the detectors, as well as factors derived from these for typical NaI(Tl) and Ge(Li) detector resolutions, are given

  8. Robustness of BW Aberrance Indices Against Test Length

    Directory of Open Access Journals (Sweden)

    Tsai-Wei Huang

    2011-09-01

    Full Text Available Many research had shown person fit indices might be influenced by the factor of test length on their detection rates of aberrant responses. The purpose of this study was to examine test length effects on the BW aberrance indices. Three conditions were designed in this study: test length (K, including 25, 50,100, and 200 items, ability ratio (T/K, defined as the total person score divided by test length K, and error ratio (E/K, defined as the number of errors within ability level divided by test length. Four 100-person times varying-item data matrices (100x25, 100x50, 100x100, and 100x200 were randomly generated and permuted 500 times for each data matrix through 20 repeats. Results showed that after partialling out the factors of E/K and T/K, the effect of test length on the association between the two indices was very slight. In nonlinear regression analyses, E/K and T/K can predict more than 76 and 73 percent of the variances of the B index and that of the W index, respectively, but test length with both very slight contributions on them. Furthermore, a very good model fit generated from SEM analyses also showed the effect of test length on the B and W indices were very tiny. All these pieces of evidence endorsed the B and W indices were robust with test length.

  9. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  10. Modification of the radiation response of pig skin by manipulation of tissue oxygen tension using anesthetics and administration of BW12C

    International Nuclear Information System (INIS)

    The importance of tissue oxygen tension on radiosensitivity was studied by examining modifications in the incidence of moist desquamation in pig skin after irradiation with strontium-90 plaques. The effects were analyzed using quantal dose-response data and comparisons were made using ED50 values for moist desquamation. Under standard anesthetic conditions of 2% halothane, approximately 70% oxygen, and approximately 30% nitrous oxide, the ED50 value (+/- SE) for moist desquamation was 27.32 +/- 0.52 Gy with no significant variation in radiosensitivity between dorsal, lateral, and ventral skin sites on the flank. Irradiation with 2% halothane and air increased the ED50 to 31.25 +/- 0.94 Gy, primarily due to an increased radioresistance of the dorsal sites. When combined with BW12C, a drug which binds oxygen selectively to hemoglobin and hence reduced the oxygen availability to tissues, a further increase in the ED50 values was observed. This was approximately 39 Gy with BW12C concentrations of 30 mg/kg and 50 mg/kg b.w. of BW12C, indicating a dose modification factor (DMF) of approximately 1.26. However, when animals were breathing the standard gas mixture, this DMF was reduced to 1.15 for 30 mg/kg of BW12C, indicating that a higher level of oxygen partly counteracted the effects of the drug in these studies with BW12C. The greatest variability in radiosensitivity was seen in the dorsal fields. This suggested complex physiological adaptation, a phenomenon that might also explain the absence of any modification of the radiation response when 100 mg/kg of BW12C was used

  11. 76 FR 23630 - Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...

    Science.gov (United States)

    2011-04-27

    ... COMMISSION Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS...: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration,...

  12. 75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

    Science.gov (United States)

    2010-11-04

    ... COMMISSION Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan Section 13.6.2, Revision 1 on Physical Security--Design Certification AGENCY: Nuclear Regulatory Commission (NRC). ACTION: Notice of Availability. SUMMARY: The NRC is issuing its Final Revision 1 to...

  13. Quality evaluation of the k0-standardized neutron activation analysis on Dalat research reactor

    International Nuclear Information System (INIS)

    Laboratory for neutron activation analysis (NAA) at the 500 kW Dalat Research Reactor( DRR) has been accredited following ISO/IEC 17025: 2005 (TCVN VILAS-519). Successful introduction of the k0-based NAA using Ko-Dalat software written in house at DRR has allowed to extend its applications in petroleum, archaeology and environment besides other traditional fields, i.e. geology, biomedicine, industry and materials. This study aimed to assess the quality of k0-NAA by analyzing a number of standard reference materials: SMELS, NIST-1547, NIST-2711a, IAEA-Soil-7 and IAEA-V-10. The laboratory has also participated in proficiency testing schemes organized by IAEA and FNCA. External and internal quality assessment revealed that the k0-NAA using Ko-Dalat software established at DRR has met the requirements of multi-element analysis in the intended applications. About 42 elements: Al, As, Au, Ba, Br, Ca, Ce, Cl, Co, Cr, Cs, Cu, Dy, Eu, Fe, Hf, I, In, K, La, Mg, Mn, Mo, Na, Nd, Pr, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, Ti, U, V, W, Yb, Zn and Zr, were determined in the above mentioned materials. The results were evaluated and reported in this paper. (author)

  14. InterProScan Result: BW999159 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available nk Molecular Function: hyaluronic acid binding (GO:0005540)|Biological Process: cell adhesion (GO:0007155) ... ...BW999159 BW999159_2_ORF1 DA92A674762AF58C PFAM PF00193 Xlink 1.8e-27 T IPR000538 Li

  15. InterProScan Result: BW999159 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available luronic acid binding (GO:0005540)|Biological Process: cell adhesion (GO:0007155) ... ...BW999159 BW999159_2_ORF1 DA92A674762AF58C PROSITE PS01241 LINK_1 NA T IPR000538 Link Molecular Function: hya

  16. InterProScan Result: BW998505 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998505 BW998505_2_ORF1 B179350D99C77D2D PRINTS PR01638 MHCCLASSI 3.9e-28 T IPR001039 MHC class ... 016020)|Biological Process: antigen processing and presentation ... (GO:0019882)|Cellular Component: MHC class I prote ...

  17. InterProScan Result: BW998505 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998505 BW998505_2_ORF1 B179350D99C77D2D PFAM PF00129 MHC_I 7.4e-72 T IPR001039 MHC class I, al ... 016020)|Biological Process: antigen processing and presentation ... (GO:0019882)|Cellular Component: MHC class I prote ...

  18. InterProScan Result: BW998505 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998505 BW998505_2_ORF1 B179350D99C77D2D GENE3D G3DSA:3.30.500.10 no description 1.4e-57 T IPR0 ... 016020)|Biological Process: antigen processing and presentation ... (GO:0019882) ...

  19. InterProScan Result: BW999554 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999554 BW999554_3_ORF1 F5F99E08C5456383 PANTHER PTHR11630 DNA REPLICATION LICENSI...NG FACTOR 1e-12 T IPR001208 unintegrated Molecular Function: DNA binding (GO:0003677)|Molecular Function: ATP binding (GO:0005524)|Biological Process: DNA replication (GO:0006260) ...

  20. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC SMART SM00017 no description 0.0087 T IPR002038 Osteopontin Biol...ogical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  1. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC PRINTS PR00216 OSTEOPONTIN 2.2e-32 T IPR002038 Osteopontin Biol...ogical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  2. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC PFAM PF00865 Osteopontin 1.9e-62 T IPR002038 Osteopontin Biol...ogical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  3. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC PANTHER PTHR10607 OSTEOPONTIN 6.2e-32 T IPR002038 Osteopontin Biol...ogical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  4. InterProScan Result: BW998975 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998975 BW998975_2_ORF1 3609DD63678CBAD1 PANTHER PTHR11461 SERINE PROTEASE INHIBITOR, SERP...IN 2.4e-109 T IPR000215 Protease inhibitor I4, serpin Molecular Function: serine-type endopeptidase inhibitor activity (GO:0004867) ...

  5. Clinical application study of bone marrow immunoscintigraphy using 99mTc-labelled anti-granulocyte monoclonal antibody BW250/183

    International Nuclear Information System (INIS)

    Objective: To study: (1) The labelling method of 99mTc-BW250/183 and its organ distribution pattern after injection. (2) The clinical value of bone marrow immunoscintigraphy of evaluation of patients with aplastic anemia. (3) The clinical value of bone marrow immunoscintigraphy of determination of bone metastasis. Methods: (1) Whole body imaging was performed to one volunteer after injection of 555 MBq 99mTc-BW250/183, meanwhile, 2 ml blood samples was taken from a cubital vein. The percentage of radioactivity of different organs and the kinetic data of in-vivo 99mTc-BW250/183 was then calculated. In all the blood samples the peripheral leukocytes were counted by a standard procedure. (2) Bone marrow immunoscintigraphy were performed to 8 patients with aplastic anemia 4 h after injection of 99mTc-BW250/183, 6 of them also underwent bone marrow imaging with 99mTc-SC. (3) Bone marrow immunoscintigraphy and conventional bone scan were performed to 14 patients with suspected bone metastases to detect bone metastases. The results was compared with X-ray, X-CT or MRI. Results: 99mTc-BW250/183 is a safe and ideal bone marrow imaging agent. Bone marrow immunoscintigraphy plays an important role in evaluating patients with aplastic anemia and determining bone metastases

  6. The Advanced Candu reactor annunciation system - Compliance with IEC standard and US NRC guidelines

    International Nuclear Information System (INIS)

    Annunciation is a key plant information system that alerts Operations staff to important changes in plant processes and systems. Operational experience at nuclear stations worldwide has shown that many annunciation implementations do not provide the support needed by Operations staff in all plant situations. To address utility needs for annunciation improvement in Candu plants, AECL in partnership with Canadian Candu utilities, undertook an annunciation improvement program in the early nineties. The outcome of the research and engineering development program was the development and simulator validation of alarm processing, display, and information presentation techniques that provide practical and effective solutions to key operational deficiencies with earlier annunciation implementations. The improved annunciation capabilities consist of a series of detection, information processing and presentation functions called the Candu Annunciation Message List System (CAMLS). The CAMLS concepts embody alarm processing, presentation and interaction techniques, and strategies and methods for annunciation system configuration to ensure improved annunciation support for all plant situations, especially in upset situations where the alarm generation rate is high. The Advanced Candu Reactor (ACR) project will employ the CAMLS annunciation concepts as the basis for primary annunciation implementations. The primary annunciation systems will be implemented from CAMLS applications hosted on AECL Advanced Control Centre Information System (ACCIS) computing technology. The ACR project has also chosen to implement main control room annunciation aspects in conformance with the following international standard and regulatory review guide for control room annunciation practice: International Electrotechnical Commission (IEC) 62241 - Main Control Room, Alarm Function and Presentation (International standard) US NRC NUREG-0700 - Human-System Interface Design Review Guidelines, Section 4

  7. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10-6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  8. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  9. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  10. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  11. The French Fast Reactor Programme: Innovations in Support of Higher Standards

    International Nuclear Information System (INIS)

    Deriving from the feedback of experience of former sodium cooled fast reactors, very high levels of requirements have been set for the ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration). Innovations are needed to further enhance safety, reduce capital cost and improve efficiency, reliability and operability, making the Generation IV SFR an attractive option for electricity production. This puts great pressure not only on the R&D and design teams at CEA but also on the CEA’s R&D and industrial partners. During the first phase of the ASTRID conceptual design (2010–2012), promising innovative options have been identified. They will be further developed in the next phases of ASTRID design studies, keeping a strong interaction between design and R&D. (author)

  12. Computer-assisted reactor NAA of geological and other reference materials, using the ksub(o)-standardization method

    International Nuclear Information System (INIS)

    USGS BCR-1 and G-2, NBS 1633a Coal Fly-Ash and a 7-element synthetic standard for biological materials were analyzed by reactor NAA, using the ksub(o)-standardization method. The analyses were performed independently in the analytical laboratories of the Institute for Nuclear Sciences (INW), Gent, and the Central Research Institute for Physics (KFKI), Budapest. This procedure allowed not only a comparison with the specified data or with other published values, but enabled a check of the consistency of our own results obtained in largely different experimental conditions. As concluded, the ksub(o)-standardization method combines general versatility (with respect to irradiation and counting conditions) with good accuracy, while the experimental work remains as simple as possible. Since the ksub(o) method is a computer-oriented technique, a FORTRAN IV program was designed and applied on a VAX 11/780 machine. (author)

  13. The effect of BW12C on radiosensitivity and necrosis of murine tissues and tumours

    International Nuclear Information System (INIS)

    BW12C is a drug that has the potential to induce normal tissue and tumour hypoxia by binding to haemoglobin, increasing its affinity for oxygen and thereby reducing oxygen availability to tissues. Initial results suggested that BW12C administration caused significant radioprotection of normal tissues and induced tumour necrosis, but variable results have been reported subsequently. This work was carried to extend the range of observations concerning the ability of BW12C to radioprotect normal tissues and tumours and to induce necrosis of tumours of the mouse. BW12C was administered as 70 mg/kg intravenous 15 min before irradiation of jejunum in CBA mice and of foot skin in WHT mice with single doses of 240 kVp X-rays while mice breathed gases of varying oxygen tensions. The radiosensitivities of these tissues were assessed by the crypt survival assay and the acute skin reaction, respectively. The radiosensitivity of CaNT tumours to single fraction irradiation was assessed by the regrowth delay assay following administration of single or multiple does of BW12C at varying times to air-breathing CBA mice. The radiation response was compared to the radiosensitivity of clamped tumours. The effect of BW12C alone on tumours was assessed by regrowth delay and histological examination for necrosis. Single or multiple doses of BW12C did not influence the radiosensitivity of CaNT tumours, although marked radioprotection could be induced by clamping the tumours during irradiation. Multiple doses of BW12C alone led to a slight increase in necrosis of the CaNT tumour but did not alter its growth rate. BW12C alone did not induce necrosis of the murine JT lymphoma. The results shown that BW12C did not have a significant effect as a radioprotective or necrotizing agent in these experimental systems. The reported differences in the radiomodifying effects of BW12C are probably tissue-specific and relate to complex biochemical and physiological interactions. 18 refs., 4 figs

  14. Study of reactor characteristics for the adaptation of the monoelement standard method in activation analysis. Application to impurity determination in silicon

    International Nuclear Information System (INIS)

    Nuclear reactions by irradiation in a nuclear reactor are reviewed. Quantitative analysis by comparison with multielement standard is treated. Comparison methods using a monoelement standard, easier to use they require a previous study of neutronic characteristics of irradiation channels of reactors (thermal and epithermal flux ratio and eventual deviations fo epithermal neutron energy spectra from the 1/E relationship). Then analysis fo silicon polycrystals by both methods is studied and interfering reactions during irradiation are examined

  15. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    International Nuclear Information System (INIS)

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  16. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hamidouche, Tewfik [Division de l' Environnement, de la Surete et des Dechets Radioactifs, Centre de Recherche Nucleaire d' Alger (CRNA), Alger (Algeria); Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: t.hamidouche@crna.dz; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)], E-mail: b.salah@ing.unipi.it; Si-Ahmed, El Khider [Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: esi-ahmed@usthb.dz; Mokeddem, Mohamed Yazid [Division de la Physique et des Applications Nucleaires, Centre de Recherche Nucleaire de Draria (CRND) (Algeria); D' Auria, Franscesco [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)

    2009-10-15

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  17. Safety-evaluation report related to the license renewal and power increase for the National Bureau of Standards Reactor (Docket No. 50-184)

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by the National Bureau of Standards (NBS) for an increase in power from 10 MWt to 20 MWt and for a renewal of the Operating License TR-5 to continue to operate the test reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Gaithersburg, Maryland, on the site of the National Bureau of Standards, which is a bureau of the Department of Commerce. The staff concludes that the NBS reactor can operate at the 20 MWt power level without endangering the health and safety of the public

  18. CFD Simulation of Hydrodynamic Characteristics in Stirred Reactors Equipped with Standard Rushton or 45°-Upward PBT Impeller

    Institute of Scientific and Technical Information of China (English)

    未作君; 徐世民; 元英进; 许松林

    2003-01-01

    The hydrodynamic characteristics generated by the standard Rushton or 45°-upward pitched-blade-turbine (PBT) impellers in a baffled reactor are numerically simulated for different off-bottom clearances (C= 1/3H and 1/2H) and agitator speeds (100, 150, 200, 250 and 300r·min-1) by using FLUENT code (Version 5.4). The results are compared with the experimental and simulated data in the published papers and good agreement is observed. The shapes of the profile of mean velocities seem independent to the speed of agitators under the experimental conditions (100-300r·min-1).

  19. A comprehensive study on the thermal movement of the reactor coolant system for Korea Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    The specific thermal pattern of each component of the reactor coolant system (RCS) is a basic design parameter for the establishment of the RCS arrangement and support system. Development of an effective method to analyze the global behavior of thermal movement of the RCS is required for design process. In this paper a fundamental but reliable and comprehensive method of obtaining thermal movement of the Korea Standard Nuclear Power Plant (KSNPP) is presented. Experimental results are included to illustrate the validity of the analyses

  20. Design of standardized WWER-1000 reactor power plant allowing industrialization of production

    International Nuclear Information System (INIS)

    The improvement consists in the siting of the individual units which allows streamlined construction, the assembly of power units and improved quality of construction work. To protect it against vibrations the reactor building is designed as a symmetric box-shaped reinforced concrete structure. The heaviest equipment is placed in the lowest parts, which increases stability and facilitates the solution of the problem of the interaction of foundations and structure proper. The cylindrical part of the sealed envelope of the reactor part may be assembled of large units up to 100 t in weight and another design of the envelope copula allows the assembly of basic equipment to be started 3 to 4 months earlier. These and other improvements make it possible to shorten construction time by 35 to 40%, to reduce material consumption and to increase productivity. (E.S.)

  1. The French Fast Reactor Program - Innovations in Support to Higher Standards

    International Nuclear Information System (INIS)

    • From the experience of ASTRID first phase of conceptual design studies (2010-2012), two remarks can be made: → Higher requirements in safety and operability lead to higher costs that cannot be fully recovered by advances in technology. This puts additional pressure on the next phases of the design to optimize the design and to keep the costs to the minimum. → There is a clear link between the level of safety that can be achieved and the maturity of the technology, i.e. the experience accumulated in R&D, design, construction, operation and decommissioning of past reactors. In the field of fast neutron reactors, this gives a strong advantage to the sodium technology, because strengths and weaknesses are well mastered. • Meeting the high requirements set for ASTRID and serving R&D needs of innovative options will require increased industrial and international collaboration

  2. The application and development of k0-standardization method of neutron activation analysis at Dalat research reactor

    International Nuclear Information System (INIS)

    In recent years the k0-NAA method has been applied and developed at the 500 kW Dalat research reactor, which includes the establishment of a PC database of k0-NAA-related nuclear parameters, e.g., radionuclide produced, half-lives, k0-factors, Q0, E-barr Eγ, etc; the access to the database is able by a k0-NAA software or by manual; the detection efficiency calibration of gamma spectrometers used in k0-NAA, the determination of reactor neutron spectrum parameters such as α and f factors and neutron fluxes in the irradiation channels, and the validation of the developed k0-NAA procedure by analysing some SRMs, namely Coal Fly Ash (NIST-1633b), Bovine Liver (NIST-1577b) and IAEA-Soil7. The analytical results showed the deviations between experimental and certified values were mostly less than 15% with most Z-scores lower than 2. The k0-NAA procedure established at the Dalat research reactor has been regarded as a reliable standardization method of NAA and as available for practical applications, in particularly for airborne particulate and crude oil samples. (author)

  3. Development and implementation of k0-INAA standardization at 10 MW Pakistan research reactor-1

    International Nuclear Information System (INIS)

    The k0-method has been developed for 10 MW Pakistan Research Reactor (PARR-1). It involved the full energy peak efficiency calibration of HPGe detector for different counting geometries and characterization of neutron flux at three irradiation channels. Neutron flux was characterized for thermal to epithermal flux ratio, epithermal flux shape factor, modified spectral index. Westcott's g-factor and fast flux. The method was validated by analyzing IAEA-SL1 (lake sediment) and NIST-SRM-1572 (citrus leaves). All calculations were performed in Excel. The results revealed most of the elements with good accuracy. (orig.)

  4. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Science.gov (United States)

    2013-09-30

    ..., incorporate licensing experience from previous design certification application reviews, and to inform... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY... AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan-draft section revision; request...

  5. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  6. Final Environmental Statement related to license renewal and power increase for the National Bureau of Standards Reactor: Docket No. 50-184

    International Nuclear Information System (INIS)

    This Final Environmental Statement contains an assessment of the environmental impact associated with renewal of Operating License No. TR-5 for the National Bureau of Standards (NBS) reactor for a period of 20 years at a power level of 20 MW. This reactor is located on the 576-acre NBS site near Gaithersburg in Montgomery County, Maryland, about 20 mi northwest of the center of Washington, DC. The reactor is a high-flux heavy-water-moderated, cooled and reflected test reactor, which first went critical on December 7, 1967. Though the reactor was originally designed for 20-MW operation, it has been operating for 14 years at a maximum authorized power level to 10 MW. Program demand is now great enough to warrant operation at a power level of 20 MW. No additional major changes to the physical plant are required to operate at 20 MW

  7. The studies for application of Ko-standardization method of neutron activation analysis on Dalat reactor

    International Nuclear Information System (INIS)

    The studies for application of the Ko-NAA method have been carried out in order to aim at creating a concrete procedure of the method, which is able to apply for practical samples. The results of the calibration of neutron spectrum parameters at cell 7-1, 13-2 and rotary rack on Dalat reactor, the calibration of detection efficiency of gamma spectrometers, the application and development of computer software for processing of gamma-ray spectra and for calculating of related problems (i.e. detector efficiency, neutron spectrum parameters, elemental concentration, etc.) and the quality control/assessment (QC/QA) with the analysis of the reference material IAEA-Soil7 have been presented as well as the application of the ko-NAA for practical samples in geological, archaeological, bio-medical and environmental objects has also been shown. (author)

  8. Offsite dose calculation manual guidance: Standard radiological effluent controls for boiling water reactors

    International Nuclear Information System (INIS)

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-- 01, which allows Radiological Effluent Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft form for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. 11 tabs

  9. Exit reactor Thetis/Ghent (1967-2003). A recollection of its significant contribution to NAA and its leading role in the development of the k0-standardization

    International Nuclear Information System (INIS)

    After 36 years of operation, reactor Thetis at the Institute for Nuclear Sciences (Ghent University) was decommissioned in December 2003. On this occasion, a survey is presented of the characteristics and features of Thetis, which were opening the way to its significant contribution to NAA and its leading role in the development of the k0-standardization. A summary is given, including a few specific examples, of fundamental analytical developments and practical applications based on irradiations in the reactor Thetis. (author)

  10. Safety evaluation report related to the license renewal and power increase for the National Bureau of Standards reactor (Docket No. 50-184)

    International Nuclear Information System (INIS)

    Supplement 1 to the Safety Evaluation Report (SER) related to the renewal of the operating license and for a power increase (10 MWt to 20 MWt) for the research reactor at the National Bureau of Standards (NBS) facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the review of the licensee's emergency plan, which had not been reviewed at the time the Safety Evaluation Report (NUREG-1007) was published, and the review of the NBS application by the Advisory Committee on Reactor Safeguards, which was completed subsequent to the publication of the SER

  11. The reference neutron field - a standard neutron source for neutron measurements at the research reactor IRT-2000 in Sofia

    International Nuclear Information System (INIS)

    A reference neutron field (RFN) is used as a standard neutron source (SNS) that is influenced by the changes in the reactor core due to recharging or other causes. A whole range of measurements is carried out in a full scope, to specify its characteristics precisely. The SNS comprises: 1) the RNF certificated to the neutron energy spectrum, its location in the reactor field, being a reference measure of the differential energy distribution in the neutron flux; 2) exposure monitoring tools (detectors revealing the certified physical characteristics); 3) functional measurement apparatus (revealing the spectral characteristics). The following basic metrological characteristics are given: differential neutron energy spectrum, described by F(E) [1/cm2.s.MeV], normalized by 1 in the range 3-19 MeV and the measurement error; the conventional neutron flux density and its error. The methodology of measuring the neutron flux integral density comprises the following six steps: 1) assessment of the influence of the changes in the core configuration on the stability of the RNF (estimated in six energy ranges); 2) demonstration of RNF application in reactor physics studies; 3) irradiation of two sets of activation detectors (Au, Sc and Au, Sc, S in Al and Cd shields); 4) measurement of the detector activities by calibrated gamma- and beta- spectrometric apparatus; 5) determination of the neutron field characteristics at a certain point of the RNF by the method of activating ratios; 6) the result accuracy assessment and probabilistic error limits determination with 95% upper bound frequency. The RNF neutron energy range have been measured 6 times for a period of two years. 6 refs., 8 figs. (M.A.)

  12. Simulation of primary to secondary break in a VVER-type reactor: Results of the IAEA's Third Standard Problem Exercise

    International Nuclear Information System (INIS)

    Seeking to enlarge the experimental data base for code assessment, the International Atomic Energy Agency (IAEA), in collaboration with the Central Research Institute for Physics of the Hungarian Academy of Sciences, has organized the Third Standard Problem Exercise (SPE-3) involving the simulation of a break from primary to secondary in the steam generator of the PMK-NVH experimental facility. The facility is a scaled-down model of a VVER-type reactor, and the experiment addresses the possibility of a break developing in the steam generator collector of the actual plant. This paper presents a brief description of the experimental facility and the experiment. Results of a comparison of pretest and posttest calculations performed by some of the 24 participants in the exercise are also presented. The complete report of the exercise has been published as an IAEA technical document

  13. Evaluation and standardization of neutron activation analysis according to the K0 method in the RP-10 reactor

    International Nuclear Information System (INIS)

    It has been characterized and standardized an irradiation of the RP-10 Research Nuclear Reactor for use of the K0 method of neutron activation analysis using the Hoegdahl convention; also it has been evaluate the behaviour of such method in regard to the accuracy and precision of the results obtained in the quantitative multi elemental analysis of several certified materials of reference. In order to prove that the analytical method is totally under statistical control, it has been used the Heydorn method. It has been verified that the method is exact, precise and reliable to determine the aluminium, antimuonium, arsenic, bromine, calcium, chloride, copper, magnesium, manganese, sodium, titanium, vanadium, zinc and other elements. Also, they are discussed, in regard to the use of K0 constants, the different formalisms employed to calculate the integral of the reaction rate by nucleus in the activation. (author). 58 refs., 18 tabs., 6 figs

  14. Standard for prevention of gas entrainment phenomena in fast reactors. (2) Proposal of gas entrainment evaluation method

    International Nuclear Information System (INIS)

    For the stable operation of fast breeder reactors (FBRs), the occurrences of gas entrainment (GE) phenomena should be suppressed below an allowance level. Therefore, a reliable evaluation method for the GE phenomena is necessary to determine the operating conditions of FBRs. However, such a method has not yet been established, especially for the vortex-type GE. In this paper, the authors propose a GE evaluation method in which free surface vortices are identified from velocity fields by using the second invariant of the velocity gradient tensor, and GE evaluation parameters, e.g., gas core length, are calculated by using the Burgers vortex model. In addition, the standard method for the prevention of three kinds of vortex-type GE is shown by considering experimental data, evaluation results obtained by the proposed method and comparison results. Finally, it is confirmed that the onset conditions of the vortex-type GE can be evaluated by the proposed method. (author)

  15. Offsite dose calculation manual guidance: Standard radiological effluent controls for pressurized water reactors

    International Nuclear Information System (INIS)

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-01, which allows Radiological Effect Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft from (NUREG-0471 and -0473) for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. Also included for completeness are: (1) radiological environmental monitoring program guidance previously which had been available as a Branch Technical Position (Rev. 1, November 1979); (2) existing ODCM guidance; and (3) a reproduction of generic Letter 89-01

  16. The protein crystallography beamline BW6 at DORIS - automatic operation and high-throughput data collection

    International Nuclear Information System (INIS)

    The wiggler beamline BW6 at DORIS has been optimized for de-novo solution of protein structures on the basis of MAD phasing. Facilities for automatic data collection, rapid data transfer and storage, and online processing have been developed which provide adequate conditions for high-throughput applications, e.g., in structural genomics

  17. Standard reference material certification: contribution of NAA with a TRIGA reactor

    International Nuclear Information System (INIS)

    Pavia has cooperative links with the major international agencies devoted to the certification of SRMs or CRMs as the Bureau Communautaire de Reference (BCR), the European Institute for Reference Materials and Measurement (IRMM), the USA National Institute of Standards and Technology (NIST) and the International Atomic Energy Agency (IAEA). During these cooperative works, a large amount of analytical data obtained with NAA has been compared, and meaningful methodological information achieved with respect to accuracy and precision in the analysis of several elements at different concentrations in various matrices. Analytical data on As, Cd, Cr, Co, Cu, Cs, Fe, Zn, K, Sc, U, Th, Al, Sb, Mn, V, Hg, Sr, Rb, Se,Pt, all the Rare Earths and halogens Br, Cl, I, have been obtained and contributed for the final certification

  18. Ageing management at the NPPs of EnBW in Germany

    International Nuclear Information System (INIS)

    This report gives a short introduction to the Ageing Management Programm (AMP) implemented at the NPPs of EnBW. The main objective of the EnKK AMP is to ensure the safe and reliable future utilizability of the four NPPs in service. The program is a relevant element for plant safety as a precaution to prevent damage and to gain a sustainable basis for a prolongation of durability. (author)

  19. BW 3200 CMTS有线调制解调器头端系统

    Institute of Scientific and Technical Information of China (English)

    2004-01-01

    太力阳BW 3200 CMTS适用天高密度的MDU以及那些为客户提供宽带接入酒店。“Pizza盒”式的CMTS是业内唯一可扩展的解决方案,它由一个单独1RU交换控制单元组成。该交换控制单元支持

  20. Untersuchung eines thermischen Lichtbogentriebwerksystems für die Lunar Mission BW1

    OpenAIRE

    Bock, Dagmar

    2009-01-01

    Die Verwendung von thermischen Lichtbogentriebwerksystemen in der 1 bis 2 kW-Klasse mit Hydrazin als Treibstoff zur Lageregelung auf geostationären Satelliten ist heutzutage Stand der Technik. In der vorliegenden Dissertation wird ein thermisches Lichtbogentriebwerksystem mit Ammoniak als Treibstoff für die Anwendung als eines der Haupttriebwerke der universitären Kleinsatellitenmission Lunar Mission BW1 des Instituts für Raumfahrtsysteme der Universität Stuttgart untersucht und charakteri...

  1. Lightcurve Analysis of the Near-Earth Asteroid (6053) 1993 BW3

    Science.gov (United States)

    Warner, Brian D.

    2015-10-01

    CCD photometric observations of the near-Earth asteroid (6053) 1993 BW3 were made at the Center for Solar System Studies-Palmer Divide Station (CS3-PDS) in 2015 January and March. Analysis of the individual and combined data sets produced a period on the order of 2.8 hours. This differs significantly from the results of Pravec et al. (1997; P = 2.573 h) and from shape models by Durech (2002) and Kaasalainen (2002). While this discordance is not resolved, the 2.573 h value has the greatest amount of data supporting it, and for now, remains the favored period solution.

  2. Study of internal rotation in molecules using molecular orbital method in the CNDO/BW approximation

    International Nuclear Information System (INIS)

    It is presented a LCAO-MO-SCF study of Internal Rotation for the molecules C2H6, CH3NH2, H2O2, and N2H4 by ysing the CNDO/BW approximation and an M-center energy partition. Our results are compared with those obtained with the CNDO/2 approximation. It is shown that there are differences in the analysis of the process involved in the internal rotation barriers mechanism. Thus the interpretation of the results is strongly dependent on the parametrization used. (author)

  3. Low-power ultraviolet lidar for standoff detection of BW agents

    Science.gov (United States)

    Prasad, Coorg R.; Huang, Wen; Bufton, Jack; Achey, Alexander; Dawson, Jeffrey; Serino, Robert M.; Shi, Wenhui

    2004-08-01

    A compact ultraviolet lidar stand-off sensor was recently developed and field-tested for detection of bio warfare (BW) agent stimulant aerosols and interferents. It employed a low-power (~5mW), continuous-wave, 375nm semiconductor ultraviolet optical source (SUVOS) laser diode that was modulated at high-speed with a pseudo-random (PR) code to provide range-resolved lidar detection of both aerosol elastic scattering and fluorescence. The sensor incorporated a 150mm diameter receiver telescope and 3 photon-counting detection channels centered at 375nm, 440nm, and 550nm. Aerosol elastic and fluorescence lidar profiles were obtained by correlating the signal photon-counts with the PR code. Tests of the lidar were performed first with simulants released in the Large Aerosol Chamber at Edgewood Chemical and Biological Center, MD at a lidar range of only 7.5m. The second phase of testing was done at Dugway Proving Ground, UT. Here the lidar was continuously scanned (+/- 13°) in a horizontal plane to detect downwind simulant and interferent aerosol disseminations at ranges of several hundred meters. Preliminary analyses of these tests show that the lidar detected fluorescence from the BW simulants at ranges up to 100m, and elastic scattering from aerosols up to 350m.

  4. SU-E-I-62: Reduction of Susceptibility Artifacts by Increasing the Bandwidth (BW) and Echo Train Length (ETL)

    Energy Technology Data Exchange (ETDEWEB)

    Mavroidis, P [University of North Carolina, Chapel Hill, NC (United States); Boci, N; Kostopoulos, S; Ninos, C; Glotsos, D; Oikonomou, G; Bakas, A; Roka, V; Cavouras, D; Lavdas, E [Technological Education Institute of Athens, Athens, Attika (Greece); Sakkas, G; Tsagkalis, A [Animus Kyanoys Larissas Hospital, Larissa, Thessaly (Greece); Chatzivasileiou, V; Batsikas, G [IASO Thessalias Hospital, Larissa, Thessaly (Greece); Papanikolaou, N [University of Texas HSC SA, San Antonio, TX (United States); Stathakis, S [Cancer Therapy and Research Center, San Antonio, TX (United States)

    2015-06-15

    Purpose: The aim of this present study is to increase bandwidth (BW) and echo train length (ETL) in Proton Density Turbo Spin Echo (PD TSE) sequences with and without fat saturation (FS) as well as in Turbo Inversion Recovery Magnitude sequences (TIRM) in order to assess whether these sequences are capable of reducing susceptibility artifacts. Methods: We compared 1) TIRM coronal (COR) with the same sequence with increased both BW and ETL 2) Conventional PD TSE sagittal (SAG) with FS with an increased BW 3) Conventional PD TSE SAG without FS with an increased BW 4) Conventional PD TSE SAG without FS with increased both BW and ETL. A quantitative analysis was performed to measure the extent of the susceptibility artifacts. Furthermore, a qualitative analysis was performed by two radiologists in order to evaluate the susceptibility artifacts, image distortion and fat suppression. The depiction of cartilage, menisci, muscles, tendons and bone marrow were also qualitatively analyzed. Results: The quantitative analysis found that the modified TIRM sequence is significantly superior to the conventional one regarding the extent of the susceptibility artifacts. In the qualitative analysis, the modified TIRM sequence was superior to the corresponding conventional one in eight characteristics out of ten that were analyzed. The modified PD TSE with FS was superior to the corresponding conventional one regarding the susceptibility artifacts, image distortion and depiction of bone marrow and cartilage while achieving effective fat saturation. The modified PD TSE sequence without FS with a high (H) BW was found to be superior corresponding to the conventional one in the case of cartilage. Conclusion: Consequently, TIRM sequence with an increased BW and ETL is proposed for producing images of high quality and modified PD TSE with H BW for smaller metals, especially when FS is used.

  5. SU-E-I-62: Reduction of Susceptibility Artifacts by Increasing the Bandwidth (BW) and Echo Train Length (ETL)

    International Nuclear Information System (INIS)

    Purpose: The aim of this present study is to increase bandwidth (BW) and echo train length (ETL) in Proton Density Turbo Spin Echo (PD TSE) sequences with and without fat saturation (FS) as well as in Turbo Inversion Recovery Magnitude sequences (TIRM) in order to assess whether these sequences are capable of reducing susceptibility artifacts. Methods: We compared 1) TIRM coronal (COR) with the same sequence with increased both BW and ETL 2) Conventional PD TSE sagittal (SAG) with FS with an increased BW 3) Conventional PD TSE SAG without FS with an increased BW 4) Conventional PD TSE SAG without FS with increased both BW and ETL. A quantitative analysis was performed to measure the extent of the susceptibility artifacts. Furthermore, a qualitative analysis was performed by two radiologists in order to evaluate the susceptibility artifacts, image distortion and fat suppression. The depiction of cartilage, menisci, muscles, tendons and bone marrow were also qualitatively analyzed. Results: The quantitative analysis found that the modified TIRM sequence is significantly superior to the conventional one regarding the extent of the susceptibility artifacts. In the qualitative analysis, the modified TIRM sequence was superior to the corresponding conventional one in eight characteristics out of ten that were analyzed. The modified PD TSE with FS was superior to the corresponding conventional one regarding the susceptibility artifacts, image distortion and depiction of bone marrow and cartilage while achieving effective fat saturation. The modified PD TSE sequence without FS with a high (H) BW was found to be superior corresponding to the conventional one in the case of cartilage. Conclusion: Consequently, TIRM sequence with an increased BW and ETL is proposed for producing images of high quality and modified PD TSE with H BW for smaller metals, especially when FS is used

  6. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  7. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Science.gov (United States)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  8. Validation of standardized computer analyses for licensing evaluation/TRITON two-dimensional and three-dimensional models for light water reactor fuel

    International Nuclear Information System (INIS)

    The isotopic depletion capabilities of the new Standardized Computer Analyses for Licensing Evaluation control module TRITON, coupled with ORIGEN-S, were evaluated using spent fuel assays from several commercial light water reactors with both standard and mixed-oxide fuel assemblies. Calculations were performed using the functional modules NEWT and KENO-VI. NEWT is a two-dimensional, arbitrary-geometry, discrete-ordinates transport code, and KENO-VI is a three-dimensional Monte Carlo transport code capable of handling complex three-dimensional geometries. To validate the codes and data used in depletion calculations, numerical predictions were compared with experimental measurements for a total of 29 samples taken from the Calvert Cliffs, Obrigheim, and San Onofre pressurized water reactors and the Gundremmingen boiling water reactor. Similar comparisons have previously been performed at the Oak Ridge National Laboratory for the one-dimensional SAS2H control module. The SAS2H, TRITON/KENO-VI, and TRITON/NEWT results were compared for corresponding samples. All analyses showed that TRITON/KENO-VI and TRITON/NEWT produced typically similar or better results than SAS2H. The calculations performed in this validation study demonstrate that the depletion capabilities of TRITON accurately model spent fuel depletion and decay. (authors)

  9. Biosimmer: A Virtual Reality Simulator for Training First Responders in a BW Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Shawver, D.M.; Sobel, A.L.; Stansfield, S.A.

    1998-11-11

    BioSimMER (Bioterrorism Simulated Medical Emergency Response) is a Virtual Reality-based mission rehearsal and training environment. BioSimMER employs contingency-oriented, multiple-path algorithms and MOESINIOPS focused on real-world operations. BioSimMER is network-based and immerses multiple trainees in a high resolution synthetic environment, including virtual casualties and instruments that they may interact with and manipulate. Trainees are represented as individuals by virtual human Avatars. The simulation consists of several components: virtual casualties dynamically manifest the symptoms of their injuries and respond to the intervention of the trainees. Agent transport analysis is used to simulate casualty exposures and to drive the responses of simulated sensors/detectors. The selected prototype scenario is representative of combined injuries anticipated in BW operations.

  10. An explicit solution to the matrix equation AV+BW=EV J

    Institute of Scientific and Technical Information of China (English)

    Aiguo WU; Guangren DUAN; Bin ZHOU

    2007-01-01

    In this note,the matrix equation AV+BW=EV J is considered,where E,A and B are given matrices of appropriate dimensions,J is an arbitrarily given Jordan matrix,V and W are the matrices to be determined.Firstly,a right factorization of(sE-A)-1 B is given based on the Leverriver algorithm for descriptor systems.Then based on this factorization and a proposed parametric solution,an alternative parametric solution to this matrix equation is established in terms of the R-controllability matrix of(E,A,B),the generalized symmetric operator and the observability matrix associated with the Jordan matrix J and a free parameter matrix.The proposed results provide great convenience for many analysis and design problems.Moreover,some equivalent forms are proposed.A numerical example is employed to illustrate the effect of the proposed approach.

  11. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  12. Effects of the 5-HT2B receptor agonist, BW 723C86, on three rat models of anxiety.

    Science.gov (United States)

    Kennett, G A; Bright, F; Trail, B; Baxter, G S; Blackburn, T P

    1996-04-01

    1. BW 723C86 (3 and 10 mg kg-1, s.c. 30 min pretest), a 5-HT2B receptor agonist, increased total interaction, but not locomotion in a rat social interaction test, a profile consistent with anxiolysis. 2. The effect of BW 723C86 in the social interaction test is likely to be 5-HT2B receptor-mediated as it was prevented by pretreatment with the 5-HT2C/2B receptor antagonist, SB 200646A, (1 and 2 mg kg-1, p.o., 1 h pretest) which did not affect basal levels of social interaction at the doses used. 3. An anxiolytic-like action was also observed in the rat Geller-Seifter conflict test, where BW 723C86 (0.5-50 mg kg-1, s.c. 30 min pretest) modestly, but significantly increased punished, but not unpublished responding. 4. In a rat 5 min elevated x-maze test, BW 723C86 (1-10 mg kg-1, s.c.) had no significant effect. 5. The maximal anxiolytic-like effect of BW 723C86 approached that of the benzodiazepine anxiolytic, chloradiazepoxide (5 mg kg-1, s.c. 30 min pretest) in the social interaction test, but was markedly less in the Geller-Siefter test. The effect of BW 723C86 was also clearly less than chlordiazepoxide in the elevated x-maze procedure where it had no significant effect. 6. In conclusion, BW 723C86 exerted an appreciable anxiolytic-like profile in a rat social interaction test, but had a weaker effect in the Geller-Siefter and was ineffective in the elevated x-maze test used. These effects are likely to be 5-HT2B receptor-mediated. PMID:8730737

  13. Determination of silicon in Japanese iron reference standard materials by reactor fast neutron activation analysis combined with a simple pre-concentration

    International Nuclear Information System (INIS)

    A reactor fast neutron activation analysis was used in combination with a simple pre-concentration procedure for determining silicon in some iron reference standard materials of Japan Iron and Steel Federation. The samples were dissolved with aqua regia and digested with perchloric or sulfuric acid. The precipitated silica was collected on a filter paper and irradiated in a cadmium case with reactor fast neutrons. Silicon can be determined in tool steel SKD6, low-alloy steel Nos 2 and 4 and silico-manganese samples by a present method measuring 1,273.4 keV γ-rays from 6.63-minute 29Al produced by 29Si(n,p)29Al reaction. (author)

  14. Method of realization and exploitation of monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors at NPP Bohunice V1

    International Nuclear Information System (INIS)

    The sequences in development of computer equipment s and the sequences in development of measurement tools and procedures are listed in submitted presentation - from start-up the power plant in operation until present days. Present status of integration of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors in NPP V1 is presented here. The ways of data acquisition, storing of results and their evaluation are described in this presentation. In conclusion some practical possibilities of using a a monitoring system for accuracy and reliability are listed. (Authors)

  15. Ductile-Phase Toughening in TiBw/Ti-Ti3Al Metallic-Intermetallic Laminate Composites

    Science.gov (United States)

    Wu, Hao; Jin, Bo Cheng; Geng, Lin; Fan, Guohua; Cui, Xiping; Huang, Meng; Hicks, Rodrigo Mier; Nutt, Steven

    2015-09-01

    The concept of ductile-phase toughening was explored in a metallic-intermetallic laminate (MIL) composite comprising alternating layers of Ti3Al and TiBw/Ti. The laminates, in which the TiBw/Ti layers were intended to impart toughness to the brittle Ti3Al, were fabricated in situ by hot pressing and reaction annealing. Compared with monolithic Ti3Al, the MIL composite exhibited marked increases in both fracture toughness and tensile elongation because of stress redistribution and strain delocalization by in situ interfaces.

  16. Spectral Response of the Pulsationally-Induced Shocks in the Atmosphere of BW Vulpeculae

    CERN Document Server

    Smith, M A; Smith, Myron A.

    2003-01-01

    The star BW Vul excites an extremely strong radial pulsation that grows in its envelope and is responsible for visible shock features in the continuum flux and spectral line profiles emerging in the atmosphere At two phases separated by 0.8 cycles. Material propelled upwards in the atmosphere from the shock returns to the lower photosphere where it creates a second shock just before the start of the next cycle. We have obtained three nights of echelle data for this star over about 5 pulsation cycles (P = 0.201 days) in order to evaluate the effects of on a number of important lines in the spectrum, including the HeI 5875A and 6678A lines. These data were supplemented by archival high-dispersion IUE (UV) data from 1994. A comparison of profiles of the two HeI lines during the peak of the infall activity suggests that differences in the development of the blue wing at this time are due to heating and short-lived formations of an optically thin layer above the atmospheric region compressed by the infall. This di...

  17. The shortest-period M-dwarf eclipsing system BW3 V38

    CERN Document Server

    Maceroni, C; Maceroni, Carla; Rucinski, Slavek M.

    1997-01-01

    The photometric data for a short-period (0.1984 day) eclipsing binary V38 discovered by the OGLE micro-lensing team in Baade's W indow field BW3 have been analyzed. The de-reddened color (V-I_C)_0=2.3 and the light-curve synthesis solution of the I-filter light curve suggest a pair of strongly-distorted M-dwarfs, with parameters between those of YY Gem and CM Dra, revolving on a tightest known orbit among binaries consisting of Main Sequence stars. The primary, more massive and hotter, component maybe filling its Roche lobe. The very small amount of angular momentum in the orbital motion makes the system particularly important for studies of angular momentum loss at the faint end of the Main Sequence. Spectroscopic observations of the orbital radial velocity variations as well as of activity indicators are urgently needed for a better understanding of the angular-momentum and internal-structure evolutionary state of the system.

  18. Development and implementation of k{sub 0}-INAA standardization at 10 MW Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Wasim, M.; Arif, M.; Zaidi, J.H.; Anwar, Y. [Chemistry Div., Pakistan Inst. of Nuclear Science and Technology, Islamabad (Pakistan)

    2009-07-01

    The k{sub 0}-method has been developed for 10 MW Pakistan Research Reactor (PARR-1). It involved the full energy peak efficiency calibration of HPGe detector for different counting geometries and characterization of neutron flux at three irradiation channels. Neutron flux was characterized for thermal to epithermal flux ratio, epithermal flux shape factor, modified spectral index. Westcott's g-factor and fast flux. The method was validated by analyzing IAEA-SL1 (lake sediment) and NIST-SRM-1572 (citrus leaves). All calculations were performed in Excel. The results revealed most of the elements with good accuracy. (orig.)

  19. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  20. Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Trac...

  1. Effects of deformation conditions on the microstructure and substructure evolution of TiBw/Ti60 composite with network structure

    Energy Technology Data Exchange (ETDEWEB)

    Wang, B.; Huang, L.J., E-mail: huanglujun@hit.edu.cn; Liu, B.X.; Geng, L.; Hu, H.T.

    2015-03-11

    The microstructure evolution of TiBw/Ti60 composite with network structure has been investigated during isothermal compression in the temperature range of 900–1100 °C and strain rate range of 0.001–1 s{sup −1}. Substructure evolution and deformation mechanism have been further investigated by electron backscatter diffraction (EBSD). In α+β phase region, the degree of flow softening decreases with the increase of temperatures and strain rates. At 900 °C, flow softening is attributed to the dynamic recovery (DRV). At 950 °C, flow softening is mainly attributed to globularization of primary α (α{sub p}) phase with the continuous dynamic recrystallization mechanism (CDRX). With decreasing strain rates, low angle grain boundaries (LAGBs) in α{sub p} phase were decreased and transformed into high angle grain boundaries (HAGBs), which can result in globularization of α{sub p} phase. In single β phase region, the prior β grain boundaries are reconstructed based on the misorientation criterion between α{sub s}-colonies using the EBSD data. Moreover, DRV, DRX and the growth of β grains occurred at low strain rates. In addition, TiBw played an important role on the microstructure evolution of matrix. At high strain rates, TiBw were seriously broken and TiBw were surrounded by lots of LAGBs. At low strain rate, DRX of β phase and globularization of α phase occurred prior near TiBw region due to providing the nucleation site for DRX and strain accumulation respectively.

  2. Effects of deformation conditions on the microstructure and substructure evolution of TiBw/Ti60 composite with network structure

    International Nuclear Information System (INIS)

    The microstructure evolution of TiBw/Ti60 composite with network structure has been investigated during isothermal compression in the temperature range of 900–1100 °C and strain rate range of 0.001–1 s−1. Substructure evolution and deformation mechanism have been further investigated by electron backscatter diffraction (EBSD). In α+β phase region, the degree of flow softening decreases with the increase of temperatures and strain rates. At 900 °C, flow softening is attributed to the dynamic recovery (DRV). At 950 °C, flow softening is mainly attributed to globularization of primary α (αp) phase with the continuous dynamic recrystallization mechanism (CDRX). With decreasing strain rates, low angle grain boundaries (LAGBs) in αp phase were decreased and transformed into high angle grain boundaries (HAGBs), which can result in globularization of αp phase. In single β phase region, the prior β grain boundaries are reconstructed based on the misorientation criterion between αs-colonies using the EBSD data. Moreover, DRV, DRX and the growth of β grains occurred at low strain rates. In addition, TiBw played an important role on the microstructure evolution of matrix. At high strain rates, TiBw were seriously broken and TiBw were surrounded by lots of LAGBs. At low strain rate, DRX of β phase and globularization of α phase occurred prior near TiBw region due to providing the nucleation site for DRX and strain accumulation respectively

  3. Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database: 1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 G...

  4. Inconsistencies of neutron flux parameters for k0 standardization in neutron activation analysis determined with the use of Au+Zr and Au+Mo+Cr monitor sets at the LVR-15 reactor in Rez

    International Nuclear Information System (INIS)

    Multipurpose research reactors require monitoring of neutron flux parameters (α, f, Fc or Φth ) during every irradiation of samples when k0 standardization in neutron activation analysis is to be used, because the above parameters may change quite frequently and unpredictably. We tested two monitor sets consisting of Au+Zr and Au+Mo+Cr for determination of the neutron flux parameters in k0-NAA at the LVR-15 reactor in Rez. Based on analysis of several reference materials, we found the use of the Au+Zr monitor set superior over the Au+Mo+Cr monitor set in the specific conditions of the LVR-15 reactor. (author)

  5. The development of a model to predict BW gain of growing cattle fed grass silage-based diets.

    Science.gov (United States)

    Huuskonen, A; Huhtanen, P

    2015-08-01

    The objective of this meta-analysis was to develop and validate empirical equations predicting BW gain (BWG) and carcass traits of growing cattle from intake and diet composition variables. The modelling was based on treatment mean data from feeding trials in growing cattle, in which the nutrient supply was manipulated by wide ranges of forage and concentrate factors. The final dataset comprised 527 diets in 116 studies. The diets were mainly based on grass silage or grass silage partly or completely replaced by whole-crop silages, hay or straw. The concentrate feeds consisted of cereal grains, fibrous by-products and protein supplements. Mixed model regression analysis with a random study effect was used to develop prediction equations for BWG and carcass traits. The best-fit models included linear and quadratic effects of metabolisable energy (ME) intake per metabolic BW (BW0.75), linear effects of BW0.75, and dietary concentrations of NDF, fat and feed metabolisable protein (MP) as significant variables. Although diet variables had significant effects on BWG, their contribution to improve the model predictions compared with ME intake models was small. Feed MP rather than total MP was included in the final model, since it is less correlated to dietary ME concentration than total MP. None of the quadratic terms of feed variables was significant (P>0.10) when included in the final models. Further, additional feed variables (e.g. silage fermentation products, forage digestibility) did not have significant effects on BWG. For carcass traits, increased ME intake (ME/BW0.75) improved both dressing proportion (P0.10) effect on dressing proportion or carcass conformation score, but it increased (P<0.01) carcass fat score. The current study demonstrated that ME intake per BW0.75 was clearly the most important variable explaining the BWG response in growing cattle. The effect of increased ME supply displayed diminishing responses that could be associated with increased

  6. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system

    International Nuclear Information System (INIS)

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO2 and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs

  7. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  8. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  9. Neutron flux characterization of the Moroccan Triga Mark II research reactor and validation of the k0 standardization method of NAA using k0-IAEA program

    International Nuclear Information System (INIS)

    The aim of this work was to implement and to validate the k0 standardization method in neutron activation analysis (k0-NAA) at the Moroccan TRIGA Mark II research reactor. This technique was used in order to determine, the calibration of several HPGe detectors and calibration of neutron flux parameters in the typical irradiation channels [rotary specimen rack (RSR) and the pneumatic tube system (PTS) facilities]. Calibrations and calculations of k0-NAA results were carried out using the k0-IAEA program. The two parameters of neutron flux in the selected irradiation channels used for elemental concentration calculation, f (thermal-to-epithermal ratio) and α (deviation from the 1/E distribution), have been determined as well in the PTS as in the RSR facilities using the zirconium bare triple method. Results obtained for f and α in two irradiation channels show that f parameter determined in this way is different in the RSR and the PTS facilities. This can be explained by the fact that the RSR channel is situated in a graphite reflector and is relatively far from the reactor core, while the PTS is in the core. Five reference materials of different origin obtained from USGS (basalt BE-N, bauxite BX-N, biotite mica-Fe, granite GS-N) and IAEA (Soil-7) were used to evaluate the validity of this method in our laboratory by analyzing the elemental concentrations with respect to the certified values. In general, good agreement was obtained between results of this work and values in certificates of the individual reference materials, thus proving the accuracy of our results and successful implementation of the method for analysis of real samples. (author)

  10. Three protagonists in B.W. Vilakazi’s “Ezinkomponi” (“On the mine compounds”)

    OpenAIRE

    N. Zondi

    2011-01-01

    In this poem the great Zulu poet B.W. Vilakazi is preoccupied with the surreal scene of a gold mine compound in the 1940s Johannesburg, and reflects on the three protagonists of the drama that plays out in front of him: the miners, mine magnates and the heavy machinery, all things that drive the entire enterprise of enslaving the workers. Feelings flood his imagination: about the terrible status of the miners (with whom he identifies); what they have left behind, their dreams and the reality ...

  11. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions

  12. Immunoscintigraphy of septic loosening of knee endoprosthesis: a retrospective evaluation of the antigranulocyte antibody BW 250/183

    Energy Technology Data Exchange (ETDEWEB)

    Klett, Rigobert; Khalisi, Alexander; Puille, Maximillian; Steiner, Dagmar; Bauer, Richard [Clinic of Nuclear Medicine, University Hospital Giessen, Friedrichstrasse 25, 35385, Giessen (Germany); Kordelle, Jens [Orthopedic Clinic, University Hospital Giessen (Germany); Stahl, Ulrich [Department of Pathology, University Hospital Giessen (Germany)

    2003-11-01

    Immunoscintigraphy with the use of the antigranulocyte antibody BW 250/183 is a reliable method for detecting infection, especially in septic loosening of hip prostheses, for which purpose quantitative interpretation of the time-activity course is employed. Therefore, we retrospectively studied whether similar results could be achieved in knee prostheses. We verified 28 scintigraphic examinations in 26 patients by histology. Scintigraphy was performed during an early (4-6 h post injection) and a late phase (23-25 h post injection). Infection was diagnosed when activity around the knee prosthesis increased by more than 10% compared with bone marrow. We found a sensitivity and a negative predictive value of 100%, a specificity of 80%, a positive predictive value of 81% and an accuracy of 89%. Specificity and accuracy are lower than in the evaluation of hip prostheses; however, in comparison to other scintigraphic modalities, scintigraphy with the antigranulocyte antibody BW 250/183 is superior in excluding infection and has better specificity and accuracy. Therefore, as in the case of hip prostheses, the described methodology appears to be the scintigraphic modality of choice for knee prostheses. (orig.)

  13. The shortest period M-dwarf eclipsing system BW3 V38, II: determination of absolute elements

    CERN Document Server

    Maceroni, C

    2004-01-01

    The spectroscopic data for the short-period (0.1984 d)eclipsing binary V38, discovered by the OGLE micro-lensing team in Baade's Window field BW3, are analyzed. Radial velocity curves are derived from mid-resolution spectra obtained with EMMI-NTT at ESO - La Silla, and a simultaneous solution of the existing light curve by OGLE and of the new radial velocity curves is obtained. The system is formed by almost twin M3e dwarf components that are very close, but not yet in contact. The spectra of both dwarfs show signatures of the presence of strong chromospheres. Spectroscopy definitely confirms, therefore, what was suggested on the basis of photometry: BW3 V38 is indeed a unique system, as no other similar binary with M components and in such a tight orbit is known. Within the limits posed by the relatively large errors, due to the combined effect of system faintness and of the constraints on exposure time, the derived physical parameters seem to agree with the relations obtained from the other few known eclips...

  14. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  15. Biosynthetic Origin of the Antibiotic Pseudopyronines A and B in Pseudomonas putida BW11M1.

    Science.gov (United States)

    Bauer, Judith S; Ghequire, Maarten G K; Nett, Markus; Josten, Michaele; Sahl, Hans-Georg; De Mot, René; Gross, Harald

    2015-11-01

    Within the framework of our effort to discover new antibiotics from pseudomonads, pseudopyronines A and B were isolated from the plant-derived Pseudomonas putida BW11M1. Pseudopyronines are 3,6-dialkyl-4-hydroxy-2-pyrones and displayed high in vitro activities against several human pathogens, and in our hands also towards the plant pathogen Pseudomonas savastanoi. Here, the biosynthesis of pseudopyronine B was studied by a combination of feeding experiments with isotopically labeled precursors, genomic sequence analysis, and gene deletion experiments. The studies resulted in the deduction of all acetate units and revealed that the biosynthesis of these α-pyrones occurs with a single PpyS-homologous ketosynthase. It fuses, with some substrate flexibility, a 3-oxo-fatty acid and a further unbranched saturated fatty acid, both of medium chain-length and provided by primary metabolism. PMID:26507104

  16. Space-time reactor kinetics for heterogeneous reactor structure

    International Nuclear Information System (INIS)

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods

  17. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  18. The clinical application study of bone marrow immunoscintigraphy using 99Tcm-BW250/183 in evaluating patients with aplastic anemia

    International Nuclear Information System (INIS)

    Objective: To study the clinical value of bone marrow immunoscintigraphy for evaluation of patients with aplastic anemia. Methods: Twelve patients with aplastic anemia underwent bone marrow immunoscintigraphy using 99Tcm labelled anti-granulocyte monoclonal antibody BW250/183, 10 of them also underwent bone marrow imaging using 99Tcm-sulfur colloid (SC) 2 - 3 days later. The semiquantitative indexes of bone marrow immunoscintigraphy of the patients were compared with those of control patients. Results: Bone marrow immunoscintigraphy was superior to 99Tcm-SC bone marrow imaging. In patients with aplastic anemia, the accumulation of 99Tcm-BW250/183 in bone marrow and spleen was lower and in liver and kidney was higher than those of control patients. Nine patients were found with multiple focal accumulation in bone marrow. Conclusion: Bone marrow immunoscintigraphy with 99Tcm-BW250/183 plays an important role in evaluating patients with aplastic anemia

  19. Examination of fast-reactor fuels and FBR analytical quality-assurance standards and methods. Progress report, October 1-December 31, 1980

    International Nuclear Information System (INIS)

    This project is directed toward the examination and comparison of the effects of neutron irradiation on Liquid Metal Fast Breeder Reactor (LMFBR) Program fuel materials. Unirradiated and irradiated materials will be examined as requested by the Reference Fuels System Branch of the Division of Reactor Research and Technology (DRRT). Capabilities have been established and are being expanded for providing conventional preirradiation and postirradiation examinations. Nondestructive tests will be conducted in a hot-cell facility specifically modified for examining irradiated prototype fuel pins at a rate commensurate with schedules established by DRRT

  20. The oEA stars QY Aql, BW Del, TZ Dra, BO Her and RR Lep: Photometric analysis, frequency search and evolutionary status

    OpenAIRE

    Liakos, Alexios; Niarchos, Panagiotis

    2014-01-01

    New and complete multi-band light curves of the oEA stars QY Aql, BW Del, TZ Dra, BO Her and RR Lep were obtained and analysed with the Wilson-Devinney code. The light curves residuals were further analysed with the Fourier method in order to derive the pulsation characteristics of the oscillating components. All the reliable observed times of minimum light were used to examine orbital period irregularities. The orbital period analyses revealed secular changes for QY Aql and BW Del, while the...

  1. Molecular and functional properties of the granulocyte specific MAb BW 250/183 suited for the immunoscintigraphic localization of inflammatory processes

    International Nuclear Information System (INIS)

    The molecular characteristics of MAb BW 250/183 like high affinity and specificity for granulocytes and high epitope density on the granulocyte membrane are optimal prerequisites for in vivo binding to human granulocytes. The lack of negative, i.e. stimulatory influences on granulocyte associated physiological functions should not result in any side effects during diagnostic application. Combined with the recently developed 99mTc labelling procedure the MAb BW 250/183-99mTc conjugate is a safe, reliable and easy to handle reagent for the successful localization of inflammatory processes. (orig.)

  2. Grinding behavior and surface appearance of (TiCp+TiBw)/Ti-6Al-4V titanium matrix composites

    Institute of Scientific and Technical Information of China (English)

    Ding Wenfeng; Zhao Biao; Xu Jiuhua; Yang Changyong; Fu Yucan; Su Honghua

    2014-01-01

    (TiCp+TiBw)/Ti-6Al-4V titanium matrix composites (PTMCs) have broad application prospects in the aviation and nuclear field. However, it is a typical difficult-to-cut material due to high hardness of the reinforcements, high strength and low thermal conductivity of Ti-6Al-4V alloy matrix. Grinding experiments with vitrified CBN wheels were conducted to analyze comparatively the grinding performance of PTMCs and Ti-6Al-4V alloy. Grinding force and force ratios, specific grinding energy, grinding temperature, surface roughness, ground surface appearance were dis-cussed. The results show that the normal grinding force and the force ratios of PTMCs are much larger than that of Ti-6Al-4V alloy. Low depth of cut and high workpiece speed are generally ben-eficial to achieve the precision ground surface for PTMCs. The hard reinforcements of PTMCs are mainly removed in the ductile mode during grinding. However, the removal phenomenon of the reinforcements due to brittle fracture still exists, which contributes to the lower specific grinding energy and grinding temperature of PTMCs than Ti-6Al-4V alloy.

  3. Protest against social inequalities in B.W. Vilakazi’s poem 'Ngoba ... sewuthi' ('Because ... you now say'

    Directory of Open Access Journals (Sweden)

    N. Zondi

    2005-07-01

    Full Text Available Long before the National Party institutionalised apartheid in 1948, individuals and organisations tried to highlight the injustices of the colonial capitalist system in South Africa, but, as Lodge (1983:6 puts it, “it all ended in speeches”. This article seeks to demonstrate how Benedict Wallet Vilakazi effectively broke the silence by bringing the plight of the black masses to the attention of the world. He strongly protested against the enslavement of black labourers, especially in the gold and diamond mines, that he depicts as responsible for the human, psychological and physical destruction of the black working classes. As a self-appointed spokesperson of the oppressed, he protested against the injustices through the medium of his poetry. One of his grave concerns was the fact that black workers had been reduced to a class with no name, no rights, practically with no life and no soul. The chosen poem “Ngoba … sewuthi” (Because … you now say is thus representative of the poems in which B.W Vilakazi externalised his commitment to the well-being of the black workers, and his protest against the insensitivity of white employers.

  4. Three protagonists in B.W. Vilakazi’s “Ezinkomponi” (“On the mine compounds”

    Directory of Open Access Journals (Sweden)

    N. Zondi

    2011-06-01

    Full Text Available In this poem the great Zulu poet B.W. Vilakazi is preoccupied with the surreal scene of a gold mine compound in the 1940s Johannesburg, and reflects on the three protagonists of the drama that plays out in front of him: the miners, mine magnates and the heavy machinery, all things that drive the entire enterprise of enslaving the workers. Feelings flood his imagination: about the terrible status of the miners (with whom he identifies; what they have left behind, their dreams and the reality they battle with; the unfeeling and overwhelming spectre of industrialisation, and distant capitalist interests; and the instruments of oppression: the deafening mine machines. These three protagonists(especially the first and the third, assume human characteristics and fight to justify their respective roles in the conflict. Vilakazi’s famous protest poem becomes a cry for help in the face of destructive industrial advancement as everpresent human drama, which pits values of gold/ and money against what is more fully human and worth living for; possibly unachievable present prosperity against a vision of future happiness and fulfilment.

  5. Role of oxidative stress in inactivation of Escherichia coli BW25113 by nanoscale zero-valent iron.

    Science.gov (United States)

    Chaithawiwat, Krittanut; Vangnai, Alisa; McEvoy, John M; Pruess, Birgit; Krajangpan, Sita; Khan, Eakalak

    2016-09-15

    An Escherichia coli BW25113 wildtype strain and mutant strains lacking genes that protect against oxidative stress were examined at different growth phases for susceptibility to zero-valent iron (nZVI). Viability of cells was determined by the plate count method. All mutant strains were more susceptible than the wild type strain to nZVI; however, susceptibility differed among the mutant strains. Consistent with the role of rpoS as a global stress regulator, an rpoS gene knockout mutant exhibited the greatest susceptibility to nZVI under the majority of conditions tested (except exponential and declining phases at longer exposure time). Mutants lacking genes encoding the inducible and constitutively expressed cytosolic superoxide dismutases, sodA and sodB, respectively, were more susceptible to nZVI than a mutant lacking the gene encoding sodC, a periplasmic superoxide dismutase. This suggests that nZVI induces oxidative stress inside the cells via superoxide generation. Quantitative polymerase chain reaction was used to examine the expression of katG, a gene encoding the catalase-peroxidase enzyme, in nZVI-treated E. coli at different growth phases. Results showed that nZVI repressed the expression of katG in all but lag phases. PMID:26953142

  6. Rapid, potentially automatable, method extract biomarkers for HPLC/ESI/MS/MS to detect and identify BW agents

    Energy Technology Data Exchange (ETDEWEB)

    White, D.C. [Univ. of Tennessee, Knoxville, TN (United States). Center for Environmental Biotechnology]|[Oak Ridge National Lab., TN (United States). Environmental Science Div.; Burkhalter, R.S.; Smith, C. [Univ. of Tennessee, Knoxville, TN (United States). Center for Environmental Biotechnology; Whitaker, K.W. [Microbial Insights, Inc., Rockford, TN (United States)

    1997-12-31

    The program proposes to concentrate on the rapid recovery of signature biomarkers based on automated high-pressure, high-temperature solvent extraction (ASE) and/or supercritical fluid extraction (SFE) to produce lipids, nucleic acids and proteins sequentially concentrated and purified in minutes with yields especially from microeukaryotes, Gram-positive bacteria and spores. Lipids are extracted in higher proportions greater than classical one-phase, room temperature solvent extraction without major changes in lipid composition. High performance liquid chromatography (HPLC) with or without derivatization, electrospray ionization (ESI) and highly specific detection by mass spectrometry (MS) particularly with (MS){sup n} provides the detection, identification and because the signature lipid biomarkers are both phenotypic as well as genotypic biomarkers, insights into potential infectivity of BW agents. Feasibility has been demonstrated with detection, identification, and determination of infectious potential of Cryptosporidium parvum at the sensitivity of a single oocyst (which is unculturable in vitro) and accurate identification and prediction, pathogenicity, and drug-resistance of Mycobacteria spp.

  7. Classification and mapping of forest type using landsat TM data and B/W infrared aerial photograph

    International Nuclear Information System (INIS)

    Accurate and cost-effective classification of forest vegetation is the primary goal for forest management and utilization of forest resources. Aerial photograph and remote sensing are the most frequent and effective method in forest resources inventories. TM and MSS are the principal observing instruments on the Landsat-4 and -5 earth observing satellite. Especially TM has considerably greater spacial, spectral, and radiometric resolution power than MSS, that is, the IFOV of TM at a nadir is 30m compared to 80m for MSS. In this study, we used TM data to classify forest types and compared the result with forest type map manufactured by interpretation of B/W infrared photographs. As a result, land use types were well defined with TM data. But classifying forest types was a little difficult and indistinct. However, the spectral signatures of forest in every season and growing stages remained as problems to be solved, and also the most effective selection and combination method of bands for differentiating the spectral plots among classes

  8. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

  9. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

    International Nuclear Information System (INIS)

    NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination

  10. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  11. Operator licensing examiner standards

    International Nuclear Information System (INIS)

    The Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining licensees and applicants for reactor operator and senior reactor operator licenses at power reactor facilities pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). The Examiner Standards are intended to assist NRC examiners and facility licensees to better understand the initial and requalification examination processes and to ensure the equitable and consistent administration of examinations to all applicants. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator licensing policy changes

  12. Standard integrated head package

    International Nuclear Information System (INIS)

    An integrated head package for a standard-type nuclear reactor is described which consolidates many components and subassemblies of the upper reactor structure into a single unit which may be removed from the reactor vessel in a single lift. Included among the consolidated elements are a pressure vessel head, a cooling shroud, control rod drive mechanisms, a missile shield, a lifting rig, a hoist assembly, and a cable tray assembly. (author)

  13. Regional Fluctuation in the Functional Consequence of LINE-1 Insertion in the Mitf Gene: The Black Spotting Phenotype Arisen from the Mitfmi-bw Mouse Lacking Melanocytes.

    Directory of Open Access Journals (Sweden)

    Kazuhisa Takeda

    Full Text Available Microphthalmia-associated transcription factor (Mitf is a key regulator for differentiation of melanoblasts, precursors to melanocytes. The mouse homozygous for the black-eyed white (Mitfmi-bw allele is characterized by the white-coat color and deafness with black eyes due to the lack of melanocytes. The Mitfmi-bw allele carries LINE-1, a retrotransposable element, which results in the Mitf deficiency. Here, we have established the black spotting mouse that was spontaneously arisen from the homozygous Mitfmi-bw mouse lacking melanocytes. The black spotting mouse shows multiple black patches on the white coat, with age-related graying. Importantly, each black patch also contains hair follicles lacking melanocytes, whereas the white-coat area completely lacks melanocytes. RT-PCR analyses of the pigmented patches confirmed that the LINE-1 insertion is retained in the Mitf gene of the black spotting mouse, thereby excluding the possibility of the somatic reversion of the Mitfmi-bw allele. The immunohistochemical analysis revealed that the staining intensity for beta-catenin was noticeably lower in hair follicles lacking melanocytes of the homozygous Mitfmi-bw mouse and the black spotting mouse, compared to the control mouse. In contrast, the staining intensity for beta-catenin and cyclin D1 was higher in keratinocytes of the black spotting mouse, compared to keratinocytes of the control mouse and the Mitfmi-bw mouse. Moreover, the keratinocyte layer appears thicker in the Mitfmi-bw mouse, with the overexpression of Ki-67, a marker for cell proliferation. We also show that the presumptive black spots are formed by embryonic day 15.5. Thus, the black spotting mouse provides the unique model to explore the molecular basis for the survival and death of developing melanoblasts and melanocyte stem cells in the epidermis. These results indicate that follicular melanocytes are responsible for maintaining the epidermal homeostasis; namely, the present study

  14. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  15. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  16. Anxiolytic-like actions of BW 723C86 in the rat Vogel conflict test are 5-HT2B receptor mediated.

    Science.gov (United States)

    Kennett, G A; Trail, B; Bright, F

    1998-12-01

    The 5-HT2B receptor agonist, BW 723C86 (10, 30(mg/kg i.p. 30 min pre-test), increased the number of punishments accepted in a rat Vogel drinking conflict paradigm over 3 min, as did the benzodiazepine anxiolytics, chlordiazepoxide (2.5-10 mg/kg p.o. 1 h pre-test) and alprazolam (0.2-5 mg/kg p.o. 1 h pre-test), but not the 5-HT2C/2B receptor agonist, m-chlorophenylpiperazine (mCPP, 0.3-3 mg/kg i.p) or the 5-HT1A receptor agonist, buspirone (5-20 mg/kg p.o. 1 h pre-test). The effect of BW 723C86 was unlikely to be secondary to enhanced thirst, as BW 723C86 did not increase the time that rats with free access to water spent drinking, nor did it reduce sensitivity to shock in the apparatus. The anti-punishment effect of BW 723C86 was opposed by prior treatment with the 5-HT2/2B receptor antagonist, SB-206553 (10 and 20 mg/kg p.o. 1 h pre-test), and the selective 5-HT2B receptor antagonist, SB-215505 (1 and 3 mg/kg p.o. 1 h pre-test), but not by the selective 5-HT2C receptor antagonist, SB-242084 (5 mg/kg p.o.), or the 5-HT1A receptor antagonist, WAY 100635 (0.1 or 0.3 mg/kg s.c. 30 min pre-test). Thus, the anti-punishment action of BW 723C86 is likely to be 5-HT2B receptor mediated. This is consistent with previous reports that BW 723C86 exhibited anxiolytic-like properties in both the social interaction and Geller-Seifter conflict tests. PMID:9886683

  17. Reactor Physics

    International Nuclear Information System (INIS)

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

  18. Reactor Physics

    International Nuclear Information System (INIS)

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

  19. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  20. Design guide for category V reactors transient reactors

    International Nuclear Information System (INIS)

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems

  1. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    The safe operation of a research reactor requires that reactor personnel be fully trained and certified by the relevant authorities. Reactor operators at PUSPATI TRIGA Reactor underwent extensive training and are certified, ever since the reactor first started its operation in 1982. With the emphasis on enhancing reactor safety in recent years, reactor operator training and certification have also evolved. This paper discusses the changes that have to be implemented and the challenges encountered in developing a new training programme to be in line with the national standards. (author)

  2. Towards standardization of the dissemination measures and tritium solubility in materials of fusion reactors; Hacia la estandarizacion de las medidas de difusion y solubilidad de tritio en materiales de reactores de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Alberto, G.; Penalva, I.; Aranburu, I.; Sarrionandia-Ibarra, A.; Legarda, F.; Martinez, P. M.; Sedano, L.; Moral, N.

    2011-07-01

    The standardization of the measurements of hydrogen isotope interaction with different materials is a challenge and goal of fusion technology programs worldwide. For decades the programs have promoted the need for a reference laboratory for measurements of hydrogen transport to the evolution of fusion technology, but that goal is still pending, in contrast to the situation in other goals I+D.

  3. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  4. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  5. Electronic and magnetic properties of Sr2MoBO6 (B=W, RE, Os): Investigation of possible half metal

    Science.gov (United States)

    Zu, Ningning; Li, Rui; Li, Qinan; Wang, Jing

    2016-02-01

    The magnetic ordering temperatures of Sr2CrBO6 (B=W, Re, Os) are the top three in the class of double perovskites so far, whereas among them only Sr2CrWO6 is a half metal. In this study, by substituting Cr with Mo, Sr2MoBO6 is investigated by using the density functional theory. The calculated results indicate that all the three Mo-based compounds exhibit the half metallic nature, in particular Sr2MoOsO6 is a compensated half metal. On the other hand, Sr2MoBO6 is estimated to have at least a comparable magnetic ordering temperature with that of Sr2CrOsO6 (experimental value of 725 K). Therefore, we expect that Sr2MoBO6 (B=W, Re, Os) would be promising candidates as spintronic materials.

  6. The oEA stars QY Aql, BW Del, TZ Dra, BO Her and RR Lep: Photometric analysis, frequency search and evolutionary status

    CERN Document Server

    Liakos, Alexios

    2014-01-01

    New and complete multi-band light curves of the oEA stars QY Aql, BW Del, TZ Dra, BO Her and RR Lep were obtained and analysed with the Wilson-Devinney code. The light curves residuals were further analysed with the Fourier method in order to derive the pulsation characteristics of the oscillating components. All the reliable observed times of minimum light were used to examine orbital period irregularities. The orbital period analyses revealed secular changes for QY Aql and BW Del, while the Light-Time Effect seems to be the best explanation for the cyclic period changes in TZ Dra and BO Her. RR Lep has a rather steady orbital period. Light curve solutions provided the means to calculate the absolute parameters of the components of the systems, which subsequently were used to make an estimate of their present evolutionary status.

  7. In vitro and in vivo studies using BW12C: toxicity, haemoglobin modification and effects on the radiosensitivity of normal marrow and RIF-1 tumours in mice

    International Nuclear Information System (INIS)

    BW12C binds to haemoglobin, shifting the oxygen saturation curve to the left, and is under investigation as an inducer of tumour hypoxia. The intrinsic cellular toxicity of the drug to RIF-1 and EMT6 cells in monolayer culture was studied, and IC50 values of 100 μg ml-1 for 24h exposure and 10 μg ml-1 for 4-day exposure were measured. (author)

  8. Controllable two-scale network architecture and enhanced mechanical properties of (Ti5Si3+TiBw)/Ti6Al4V composites.

    Science.gov (United States)

    Jiao, Y; Huang, L J; Duan, T B; Wei, S L; Kaveendran, B; Geng, L

    2016-01-01

    Novel Ti6Al4V alloy matrix composites with a controllable two-scale network architecture were successfully fabricated by reaction hot pressing (RHP). TiB whiskers (TiBw) were in-situ synthesized around the Ti6Al4V matrix particles, and formed the first-scale network structure (FSNS). Ti5Si3 needles (Ti5Si3) precipitated in the β phase around the equiaxed α phase, and formed the secondary-scale network structure (SSNS). This resulted in increased deformation compatibility accompanied with enhanced mechanical properties. Apart from the reinforcement distribution and the volume fraction, the ratio between Ti5Si3 and TiBw fraction were controlled. The prepared (Ti5Si3 + TiBw)/Ti6Al4V composites showed higher tensile strength and ductility than the composites with a one-scale microstructure, and superior wear resistance over the Ti6Al4V alloy under dry sliding wear conditions at room temperature. PMID:27622992

  9. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  10. Water Cooled FBNR Nuclear Reactor

    International Nuclear Information System (INIS)

    A new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are developed by the INPRO program of the International Atomic Energy Agency (IAEA). The IAEA is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The Fixed Bed Nuclear Reactor (FBNR) is a small (70 MWe) nuclear reactor that meets all the established requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve as a dual purpose plant to produce simultaneously both electricity and desalinated water thus making it especially suitable to the needs of most of developing countries. FBNR is developed with the support of the IAEA under its program of Small Reactors Without On-Site Refuelling (SRWOSR). The FBNR reactor uses the pressurized water reactor (PWR) technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact. (author)

  11. Research reactors

    International Nuclear Information System (INIS)

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  12. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  13. Research reactors

    International Nuclear Information System (INIS)

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  14. Reactor container

    International Nuclear Information System (INIS)

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  15. Design guide for category IV reactors: liquid metal reactors

    International Nuclear Information System (INIS)

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of liquid metal cooled fast reactor (Category IV reactor) structures, components, and systems

  16. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors

    International Nuclear Information System (INIS)

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  17. Safety of research reactors

    International Nuclear Information System (INIS)

    review of the research reactor facility and to verify compliance with the IAEA' Safety Standards. This issue paper discusses the concerns generated by an analysis of the results of INSARR missions and those expressed by INSAG. The topic is timely and important because a large number of research reactors currently face not only these concerns but also the problem of spent fuel disposal following completion of the current US takeback programme and the Russian take-back programme, which is expected to commence in the near future. Many countries will need to make decisions soon on the future of their reactors to take advantage of these spent fuel options

  18. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

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

  19. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

  20. Characterization of low power research reactor neutrons for the validation of k0-INAA standardization based on k0-IAEA software

    International Nuclear Information System (INIS)

    Channel Bsite2 of Ghana research reactor-1 has been characterized for k0-INAA application. Cadmium ratio and bare multi-monitor were used to determine flux parameters using 0.1%Au–Al, Fe, and Zr wire as flux monitors. The parameters determined were 18.36±1.91, 0.0479±0.012, 5.12×1011±0.42×1011 ncm−2 s−1, 2.74×1010±0.14×1010 ncm−2 s−1, 7.73×1010±0.16×1010 ncm−2 s−1 and 16.75±1.58, −0.034±0.0028, 4.28×1011±1.71×1011 ncm−2 s−1, 2.55×1010±0.15×1010 ncm−2 s−1 respectively for thermal-to-epithermal flux ratio, alpha, thermal neutron, epithermal neutron and fast neutron flux using cadmium ratio and multi-monitor method accordingly. The k0-INAA performance assessment based on z-score distributions showed most results within |z|0-INAA application. • Both bare and Cd-ratio method were used for the characterization. • α, f and thermal, epithermal and fast neutron flux values were found to agree largely with earlier reports. • We developed an irradiation method to detect and quantify up to 32 elements in plants, rock and lake sediment matrices. • The z-score analysis showed 90% of the data points within the range of |z|<2

  1. Standard for prevention of gas entrainment phenomena in fast reactors. (1) Validations of CFD methods for reproducibilities of gas entrainment phenomena

    International Nuclear Information System (INIS)

    It is of importance for stable operations of sodium-cooled fast reactors (SFRs) to prevent gas entrainment (GE) phenomena due to free surface vortices. The entrained gas flow rate should be below an allowance level. However, theoretical determination of universal onset conditions of GE is difficult due to nonlinear characteristics of GE phenomena. Therefore, the authors have been developing an evaluation method for GE based on computational fluid dynamics (CFD) methods. In this study, we determine a suitable CFD method for GE phenomena from several candidates through some numerical benchmarks. As a result, we obtain the following guideline for the vortex-induced gas entrainment. The free vortex flow around the vortex core can be correctly evaluated by using appropriate numerical models, such as sufficient mesh resolution, suitable advection solver, suitable turbulence and free surface modeling. From the geometrical viewpoint, the jagged description of the curved boundary using a rectangular mesh is not suitable since it damps rotating flow. As for turbulence modeling, which is especially investigated in this paper, direct numerical simulation (DNS) without any turbulent model is strongly recommended, but RNG k-ε and LES are acceptable. Lastly, we apply the recommended methods to the numerical analysis of a large-scale (>1/2) test experiment. The numerical results show good agreement with the onset condition of the GE observed in the experiments. This fact indicates that our recommended CFD methods are applicable to the GE phenomena in SFRs. In the next paper, a GE evaluation method is developed, which can calculate GE occurrences based on the results of numerical simulations performed in accordance with the simulation guideline proposed in this paper. (author)

  2. New fission reactor designs

    International Nuclear Information System (INIS)

    A number of critical challenges to the expanded or continued use of nuclear power have developed. These can be categorized as: regulatory restrictions and complications; negative public attitudes; plant complexity; plant life, operations, and maintenance; uncertain load growth, financing; waste management. Solutions to these challenges through advanced reactor design centre around four key technical responses. Passive safety systems are being introduced which use the laws of physics to provide emergency reactor coding, control and shutdown thus eliminating the possibility of human error. Modular construction promises cuts in costs and construction time by shifting the major part of component manufacture from the site to the factory. Standardization also cuts capital costs and in addition operations and repair costs and expedites reactor licensing. Improvements to the fuel cycle include improved fuel types, designs and fabrication, and the reprocessing of and recycling spent fuel back into energy production, thus extending uranium resources and offering a partial solution to the problem of waste disposal. Examples of evolutionary and advanced water-cooled reactors, modular high temperature gas-cooled reactors, and advanced liquid metal cooled fast breeder reactors which are being developed round the world are presented. (author)

  3. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U3Si2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  4. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  5. Heterogeneous reactors

    International Nuclear Information System (INIS)

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author)

  6. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  7. Order-disorder in In3+ perovskites: The example of A(In2/3B''1/3)O3 (A=Ba, Sr; B''=W, U)

    International Nuclear Information System (INIS)

    We describe the preparation and structural characterization of four In-containing perovskites from neutron powder diffraction (NPD) and X-ray powder diffraction (XRPD) data. Sr3In2B''O9 and Ba(In2/3B''1/3)O3 (B''=W, U) were synthesized by standard ceramic procedures. The crystal structure of the W-containing perovskites and Ba(In2/3U1/3)O3 have been revisited based on our high-resolution NPD and XRPD data, while for the new U-containing perovskite Sr3In2UO9 the structural refinement was carried out from high-resolution XRPD data. At room temperature, the crystal structure for the two Sr phases is monoclinic, space group P21/n, where the In atoms occupy two different sites Sr2[In]2d[In1/3B''2/3]2cO6, with a=5.7548(2) A, b=5.7706(2) A, c=8.1432(3) A, β=90.01(1)o for B''=W and a=5.861(1) A, b=5.908(1) A, c=8.315(2) A, β=89.98(1)o for B''=U. The two phases with A=Ba should be described in a simple cubic perovskite unit cell (S.G. Pm3-bar m) with In and B'' distributed at random at the octahedral sites, with a=4.16111(1) A and 4.24941(1) A for W and U compounds, respectively. - Graphical abstract: The structure of the new uranium-based double perovskite Sr3In2UO9 is described and the true symmetry of the other title compounds are revisited. The presence of long-range ordering in the Sr samples, by contrast with the Ba perovskites, is related with the smaller unit cell and B-B distances in the Sr oxides, promoting the electrostatic repulsions between highly charged W6+ and U6+ cations as driving force for the long-range B-site ordering

  8. 40 CFR 63.1406 - Reactor batch process vent provisions.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions...

  9. Plasma reactor

    OpenAIRE

    Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric

    2008-01-01

    [EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.

  10. Reactor physics

    International Nuclear Information System (INIS)

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  11. ''bw-cell'': development of a natural gas powered PEM fuel cell system in the 4-kW{sub e} class; ''bw-cell'': Entwicklung eines Erdgasbetriebenen PEM-Brennstoffzellensystems in der 4-kW{sub e}-Klasse

    Energy Technology Data Exchange (ETDEWEB)

    Loeffler, M.S.; Baumgart, F.; Specht, M. [Zentrum fuer Sonnenenergie- und Wasserstoff-Forschung (ZSW), Stuttgart (Germany); Lehnert, W. [Zentrum fuer Sonnenenergie- und Wasserstoff-Forschung (ZSW), Ulm (Germany); Schmid, H.P. [WS Reformer GmbH, Renningen (Germany)

    2007-01-15

    Funded by the Environment Ministry of the federal state of Baden-Wuerttemberg, ZSW has developed a {mu}-CHP PEM fuel cell system for domestic energy supply in the 4-kW{sub e} class (''bw-cell''). The net electric efficiency is above 30% at loads from 50 to 100%. Without principal design modifications, up to 35% efficiency can be realised by optimisation of the auxiliary components and the inverter. The ''bw-cell'' weight of 200 kg with a refrigerator volume reveals the potential for future series production. Next development steps comprise the refitting to fossile/regenerative energy carriers like LPG (liquefied petroleum gas), ethanol, methanol and DME (dimethyl ether). (orig.)

  12. Broiler performance, hatching egg, and age relationships of progeny from standard and dwarf broiler dams.

    Science.gov (United States)

    Tahir, M; Cervantes, H; Farmer, C W; Shim, M Y; Pesti, G M

    2011-06-01

    The relationship of egg and chick weights to the performance of broiler chickens from two 42-wk-old flocks (standard and dwarf dams) having male parents from the same genetic stock was investigated in this study. Fertility (91.7 vs. 94.7%) and hatchability (95.2 vs. 96.3%) were not significantly (P > 0.10) different for eggs from standard and dwarf dams, respectively. Egg weight contributed significantly to the variation in BW [BW = β(0) + β(i) (egg weight) + β(i) (dam) + β(i) (sex)]. Body weight as a function of chick weight was not significant. However, chick weight was significant when included in a model with egg weight, suggesting that significant differences in BW at 50 d could be attributed to both egg and chick weights. The negative coefficient for chick weight indicated that between the 2 broilers of the same egg weight, the one with the greater chick weight would have the smaller 50-d BW. Chick weight was a linear function of egg weight. Similarly, the effect of egg or chick weight on broiler BW at 35 or 50 d was best represented by a single linear function. Dam genotype did not contribute significantly to variation in 50-d BW after variation attributable to egg weight was removed from the model. Differences in BW attributable to egg weight increased with broiler age. The coefficients of egg weight and chick weight showed that the differences in BW per gram of egg were 1.43, 3.06, 6.24, and 7.61 g and those per gram of chick were 1.87, 3.99, 8.14, and 9.93 g, respectively, at 7, 21, 35, and 50 d. Body weight increased by 0.1563 times egg weight (and 0.2092 times chick weight) with each additional day of age for both sexes and genotypes. Clearly, both egg and chick weights are important for modeling or predicting market-age broiler BW and economic returns. The relatively small relationship between BW and egg weight demonstrates that genetic selection over the past 3 decades has decreased the influence of egg weight on broiler growth. The present dwarf

  13. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

  14. China experimental fast reactor

    International Nuclear Information System (INIS)

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*1015 n/cm2/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  15. 乙肝疫苗新型佐剂 CpG-BW006对小鼠脾 NK 细胞表面分子 CD69的表达及γ干扰素分泌的影响%Influences of CpG Adjuvant BW006 of Hepatitis B Vaccine on Mu-rine Spleen NK Cells Surface CD69 Expression and IFN-γ Secretion

    Institute of Scientific and Technical Information of China (English)

    舒凯; 黄秋香; 张现臣

    2013-01-01

    Objective: To study the influences of CpG adjuvant BW006 for hepatitis B vaccine on the pheno⁃type and functions of murine spleen natural killer(NK) cells. Methods: After murine spleen NK cells were stimu⁃lated with BW006 and/or HBsAg in vitro, the expression of CD69 on NK cells surface was analyzed by flow cy⁃tometer and the secretion level of IFN-γ was detected by ELISA. Results: After 24 hours of stimulation, the ex⁃pression of CD69 reached peak and the expression rate was 48.18% which was significantly higher than 40 μg HBsAg group(21.44%, P<0.05). However, the expression of CD69 on NK cells surface have no significant differ⁃ence in 5 μg BW006 group and 5 μg BW006 combined with 40 μg HBsAg group(58.49%). The IFN-γ secre⁃tion levels of murine spleen NK cells was up to 56.95 ng/mL after activating with 5 μg BW006 which was signifi⁃cantly higher than 40 μg HBsAg group(8.47 ng/mL, P<0.05). However, the IFN-γ secretion levels of murine spleen NK cells have no significant difference in 5 μg BW006 group and 5 μg BW006 combined with 40 μg HB⁃sAg group(57.70%). Conclusion: BW006 functions to activate NK cells on early stage, regulate expression of CD69 and induce IFN-γ secretion, indicating BW006 to be a novel promising vaccine adjuvant.%  目的:研究重组乙型肝炎表面抗(HBsAg)佐剂 BW006对小鼠脾自然杀伤(NK)细胞表面分子 CD69的表达和γ干扰素(IFN-γ)分泌水平的影响.方法:BW006、HBsAg 单用或联用体外刺激小鼠脾 NK 细胞,流式细胞仪检测NK 细胞膜表面分子 CD69的表达水平,ELISA 检测 IFN-γ的分泌水平.结果:5μg BW006体外刺激小鼠脾 NK 细胞24 h 后,NK 细胞表面分子 CD69的表达达峰值(阳性率45.18%),显著高40μg HBsAg 组(21.44%)(P<0.05),与5μg BW006和40μg HBsAg 联用组(58.49%)相比无显著差异;24 h 时,5μg BW006组的 IFN-γ分泌水平达56.95 ng/mL,显著高40μg HBsAg 组(8.74 ng/mL)(P<0.05),与联用组(57.70 ng/mL)相比无显著差异.结论:BW

  16. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

    The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

  17. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  18. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

  20. The incumbent German power companies in a changing environment. A comparison of E.ON, RWE, EnBW and Vattenfall from 1998 to 2013

    Energy Technology Data Exchange (ETDEWEB)

    Kungl, Gregor

    2014-07-01

    This paper examines the actions and strategies of Germany's leading energy companies - E.ON, RWE, EnBW and Vattenfall - in the light of a changing regulatory framework and other circumstances. The liberalization of the German electricity market, measures to promote renewable energies, market developments as well as exogenous shocks such as the Fukushima nuclear disaster and the fiscal crisis all had far-reaching consequences for these companies. A comparative analysis of these companies from 1998 to 2013 shows their development from thriving growth at the start of liberalization up to the current state of crisis. Conducted with a focus on the context of the Energiewende - Germany's commitment to shift towards sustainable energy production - this article contributes to the current debate on the sustainable transformation of energy supply. The theory of strategic action fields by Fligstein and McAdam serves as a theoretical framework.

  1. The incumbent German power companies in a changing environment. A comparison of E.ON, RWE, EnBW and Vattenfall from 1998 to 2013

    International Nuclear Information System (INIS)

    This paper examines the actions and strategies of Germany's leading energy companies - E.ON, RWE, EnBW and Vattenfall - in the light of a changing regulatory framework and other circumstances. The liberalization of the German electricity market, measures to promote renewable energies, market developments as well as exogenous shocks such as the Fukushima nuclear disaster and the fiscal crisis all had far-reaching consequences for these companies. A comparative analysis of these companies from 1998 to 2013 shows their development from thriving growth at the start of liberalization up to the current state of crisis. Conducted with a focus on the context of the Energiewende - Germany's commitment to shift towards sustainable energy production - this article contributes to the current debate on the sustainable transformation of energy supply. The theory of strategic action fields by Fligstein and McAdam serves as a theoretical framework.

  2. Multifunctional reactors

    OpenAIRE

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...

  3. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  5. Breeder reactors

    International Nuclear Information System (INIS)

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components

  6. Calorimetric dosimetry of reactor radiation

    International Nuclear Information System (INIS)

    Calorimetric dosimetry of reactor radiation is relatively new reactor dosimetry method and the number of relevant papers is rather small. Some difficulties in applying standard methods (chemical dosemeters, ionization chambers) exist because of the complexity of radiation. In general application of calorimetric dosemeters for measuring absorbed doses is most precise. In addition to adequate choice of calorimetric bodies there is a possibility of determining the yields of each component of the radiation mixture in the total absorbed dose. This paper contains a short review of the basic calorimetry methods and some results of measurements at the RA reactor in Vinca performed by isothermal calorimeter

  7. Ageing management for research reactors

    International Nuclear Information System (INIS)

    During the past several years, ageing of research reactor facilities continues to be an important safety issue. Despite the efforts exerted by operating organizations and regulatory authorities worldwide to address this issue, the need for an improved strategy as well as the need for establishing and implementing a systematic approach to ageing management at research reactors was identified. This paper discusses, on the basis of the IAEA Safety Standards, the effect of ageing on the safety of research reactors and presents a proactive strategy for ageing management. A systematic approach for ageing management is developed and presented together with its key elements, along with practical examples for their application. (author)

  8. NDE standards for materials of fast reactors

    International Nuclear Information System (INIS)

    Primary objective of this paper is to bring out the salient features of the specifications followed for the procurement of various materials such as 316LN plates, tube sheets and 9Cr-1Mo tube sheets, 2.25Cr-1Mo dished ends and chrome-moly tubes and the difficulties encountered in procurement. 4 figs

  9. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  10. Reactor utilization

    International Nuclear Information System (INIS)

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  11. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  12. The AP1000 reactor

    International Nuclear Information System (INIS)

    The design of the AP1000 reactor began 20 years ago when Westinghouse launched the AP600 reactor project. In fact by re-assessing AP600's safety margins Westinghouse realized that the its power output could be raised without putting at risk its safety standard. The AP1000 was born, it yields 1100 MWe. The main AP1000's design features is its passive safety (particularly after the Fukushima accident) and its modularity. The passive safety of the AP1000 implies: -) no humane intervention needed for 72 hours at least after the incident; -) no necessity for redundant complex safety systems. The modularity means that the plant, the reactor and other buildings are constructed from a choice of 300 modular units. These units can be built off-site and fit together on site. The modularity allows more construction activities to be led simultaneously and more chances to cope with the construction schedule. The NRC has approved the operation license for 30 years of the first AP1000 being built in the Usa (Vogtle plant in Georgia). 4 AP1000 are being built in China (Sanmen and Haiyang sites) and 6 others are planned in the Usa. Westinghouse is convinced that the AP1000's passive safety makes it more attractive. Let us not forget that Westinghouse was at the origin of the concept of pressurized water reactors, an idea adopted for half the nuclear power stations in the world and for all the plants now active in France. (A.C.)

  13. Elements of reactor system design

    International Nuclear Information System (INIS)

    When the first commercial nuclear power plants were designed, each plant was treated as a new design problem. However, it became apparent that the full design effort was far too lengthy and costly to be undertaken for each order. The reactor system vendors have therefore developed a series of essentially standard reactor designs. A utility customer is offered that standard design which most closely meets his requirements. Only minor modification are made in order to meet particular local requirements. The reactor design effort for such a plant is generally limited to (a) a verification that the standard system proposed will meet the required specifications and (b) a revision of the safety analysis to take into consideration the features of the particular site. Standard system designs are usually revised on a regular basis to take advantage of new developments and operational experience. It has become customary to refer to the reactor core and entire primary system as the ''nuclear steam supply system''. In the United States, when a reactor vendor supplies a system to a public utility, it is generally only the ''nuclear steam supply system'' and specific auxiliaries which are supplied. The reactor vendor will specify the general requirements of the steam cycle, vapor container and auxiliary systems and safety systems which are not vendor supplied. The detailed design of these systems, as well as the complete structural and electrical design, is normally handled by the utility or an architect-engineer engaged by the utility. The safety analysis is usually conducted by the reactor vendor. As more experience with nuclear systems is gained, it is likely that the larger utilities will assume an expanded role in the design process

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  15. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  16. Russian-American venture designs new reactor

    International Nuclear Information System (INIS)

    Russian and American nuclear energy experts have completed a joint design study of a small, low-cost and demonstrably accident-proof reactor that they say could revolutionize the way conventional reactors are designed, marketed and operated. The joint design is helium-cooled and graphite-moderated and has a power density of 3 MWt/cubic meter, which is significantly less than the standard American reactor. A prototype of this design should be operating in Chelyabinsk by June 1996

  17. Physics and kinetics of TRIGA reactor

    International Nuclear Information System (INIS)

    This training module is written as an introduction to reactor physics for reactor operators. It assumes the reader has a basic, fundamental knowledge of physics, materials and mathematics. The objective is to provide enough reactor theory knowledge to safely operate a typical research reactor. At this level, it does not necessarily provide enough information to evaluate the safety aspects of experiment or non-standard operation reviews. The material provides a survey of basic reactor physics and kinetics of TRIGA type reactors. Subjects such as the multiplication factor, reactivity, temperature coefficients, poisoning, delayed neutrons and criticality are discussed in such a manner that even someone not familiar with reactor physics and kinetics can easily follow. A minimum of equations are used and several tables and graphs illustrate the text. (author)

  18. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

  19. Applications of Research Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

  20. Application of BW Water Base Quenching Liquid in Aluminum-ally Sheet Metal Component Quenching%BW水基淬火液在铝合金钣金件淬火中的应用

    Institute of Scientific and Technical Information of China (English)

    绳小龙

    2012-01-01

    The research of application of BW water base quenching liquid in aluminum-ally was complicated,the aspects such as mechanical property,quenching deformation,intercrystalline corrosion when BW water base quenching liquid was did to aluminum-alloy were discussed.%介绍了BW水基淬火液在铝合金钣金件中的应用研究,对铝舍金采用BW水基淬火液淬火并时效后的力学性能、淬火变形、晶间腐蚀等方面进行了探讨.

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  2. Neutronic models for the HIFAR reactor

    International Nuclear Information System (INIS)

    Standard neutronic models have been developed for the AAEC's materials testing reactor HIFAR, and are available as members of a partitioned data set. The models have been used to calculate reactor physics parameters related to operation and safety. Results from the calculations are presented

  3. Selecting reactor operator trainees

    International Nuclear Information System (INIS)

    Reactor operator trainee selection tends to be more effective if tailored to a utility's unique needs, and offers the organization a better chance for compliance with Federal regulations than if selection methods are adopted without benefit of local research. The costs of operator training range from $50,000 to $100,000. The test validity relative to a variety of training grades and performance measures is reviewed. Of interest is the degree to which tests differentiate reactor operators with respect to simulator training grades and performance in simulator operation; forms of evaluation which have become fairly standard throughout the power industry. The tests administered to each individual were selected because of their presumed relevance to training grades, and the aptitude measures are intended to assess an individual's potential to benefit from training. Tests, availability, form, the abilities they measure, and the time limit are described. (MCW)

  4. The properties of the host galaxy and the immediate environment of GRB 980425 / SN 1998bw from the multi-wavelength spectral energy distribution

    CERN Document Server

    Michałowski, Michał J; Malesani, Daniele; Michałowski, Tadeusz; Cerón, José María Castro; Reinfrank, Robert F; Garrett, Michael A; Fynbo, Johan P U; Watson, Darach J; Jørgensen, Uffe G

    2008-01-01

    We present an analysis of the spectral energy distribution (SED) of the galaxy ESO 184-G82, the host of the closest known long gamma-ray burst (GRB) 980425 and its associated supernova SN 1998bw. We use our observations obtained at the Australia Telescope Compact Array (the third >3 sigma radio detection of a GRB host) as well as archival infrared (IR) and ultraviolet (UV) observations to estimate its star formation state. We find that ESO 184-G82 has a UV star formation rate (SFR) and stellar mass consistent with the population of cosmological GRB hosts and of local dwarf galaxies. It has however a higher specific SFR (per unit stellar mass) and lower molecular gas-to-dust ratio than luminous spiral galaxies. The mass of ESO 184-G82 is dominated by an older stellar population in contrast to the majority of GRB hosts. The Wolf-Rayet region ~800 pc from the supernova site experienced a starburst episode during which the majority of its stellar population was built up. Unlike that of the entire galaxy, its SED ...

  5. Grinding behavior and surface appearance of (TiCp + TiBw/Ti-6Al-4V titanium matrix composites

    Directory of Open Access Journals (Sweden)

    Ding Wenfeng

    2014-10-01

    Full Text Available (TiCp + TiBw/Ti-6Al-4V titanium matrix composites (PTMCs have broad application prospects in the aviation and nuclear field. However, it is a typical difficult-to-cut material due to high hardness of the reinforcements, high strength and low thermal conductivity of Ti-6Al-4V alloy matrix. Grinding experiments with vitrified CBN wheels were conducted to analyze comparatively the grinding performance of PTMCs and Ti-6Al-4V alloy. Grinding force and force ratios, specific grinding energy, grinding temperature, surface roughness, ground surface appearance were discussed. The results show that the normal grinding force and the force ratios of PTMCs are much larger than that of Ti-6Al-4V alloy. Low depth of cut and high workpiece speed are generally beneficial to achieve the precision ground surface for PTMCs. The hard reinforcements of PTMCs are mainly removed in the ductile mode during grinding. However, the removal phenomenon of the reinforcements due to brittle fracture still exists, which contributes to the lower specific grinding energy and grinding temperature of PTMCs than Ti-6Al-4V alloy.

  6. A decision-tree model to detect post-calving diseases based on rumination, activity, milk yield, BW and voluntary visits to the milking robot.

    Science.gov (United States)

    Steensels, M; Antler, A; Bahr, C; Berckmans, D; Maltz, E; Halachmi, I

    2016-09-01

    Early detection of post-calving health problems is critical for dairy operations. Separating sick cows from the herd is important, especially in robotic-milking dairy farms, where searching for a sick cow can disturb the other cows' routine. The objectives of this study were to develop and apply a behaviour- and performance-based health-detection model to post-calving cows in a robotic-milking dairy farm, with the aim of detecting sick cows based on available commercial sensors. The study was conducted in an Israeli robotic-milking dairy farm with 250 Israeli-Holstein cows. All cows were equipped with rumination- and neck-activity sensors. Milk yield, visits to the milking robot and BW were recorded in the milking robot. A decision-tree model was developed on a calibration data set (historical data of the 10 months before the study) and was validated on the new data set. The decision model generated a probability of being sick for each cow. The model was applied once a week just before the veterinarian performed the weekly routine post-calving health check. The veterinarian's diagnosis served as a binary reference for the model (healthy-sick). The overall accuracy of the model was 78%, with a specificity of 87% and a sensitivity of 69%, suggesting its practical value. PMID:27221983

  7. Sodium fast neutron reactors. Status and perspective of development

    International Nuclear Information System (INIS)

    This report reveals data on development history of domestic fast neutron reactors cooled with sodium (BN reactors). It also shows BN reactors' unique role in expanding source of nuclear power raw materials and in solving ecological problems relating to radioactive wastes. There is brief information on characteristics and operation experience of research reactors BR-10, BOR-60, pilot-industrial reactors BN-350 and BN-600. As well there is data on BN-800 reactor designing that obtained a license for building. There are considered BN reactor peculiarities in regard of safety and design decisions on safety provision at the level meeting standard document requirements. BN reactor technical and economic indices and the ways of their improvement are evaluated. There is brief information on alternative perspective technologies of fast reactors, in particular regarding 'BREST-300' reactor cooled with lead coolant

  8. Nuclear reactor fissile isotopes antineutrino spectra

    OpenAIRE

    Sinev, V.

    2012-01-01

    Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using s...

  9. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  10. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  11. Reactor building

    International Nuclear Information System (INIS)

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  13. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  16. Standard interface file handbook

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, A.; Huria, H.C. [Cincinnati Univ., OH (United States)

    1992-10-01

    This handbook documents many of the standard interface file formats that have been adopted by the US Department of Energy to facilitate communications between and portability of, various large reactor physics and radiation transport software packages. The emphasis is on those files needed for use of the VENTURE/PC diffusion-depletion code system. File structures, contents and some practical advice on use of the various files are provided.

  17. Standard interface file handbook

    International Nuclear Information System (INIS)

    This handbook documents many of the standard interface file formats that have been adopted by the US Department of Energy to facilitate communications between and portability of, various large reactor physics and radiation transport software packages. The emphasis is on those files needed for use of the VENTURE/PC diffusion-depletion code system. File structures, contents and some practical advice on use of the various files are provided

  18. Simplified numerical simulation of hot channel in sodium cooled reactor

    International Nuclear Information System (INIS)

    The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author)

  19. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  20. Operator licensing examiner standards

    International Nuclear Information System (INIS)

    The Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining licensees and applicants for reactor operator and senior reactor operator licenses at power reactor facilities pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). The Examiner Standards are intended to assist NRC examiners and facility licensees to better understand the initial and requalification examination processes and to ensure the equitable and consistent administration of examinations to all applicants. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator licensing policy changes. Revision 7 was published in January 1993 and became effective in August 1993. Supplement 1 is being issued primarily to implement administrative changes to the requalification examination program resulting from the amendment to 10 CFR 55 that eliminated the requirement for every licensed operator to pass an NRC-conducted requalification examination as a condition for license renewal. The supplement does not substantially alter either the initial or requalification examination processes and will become effective 30 days after its publication is noticed in the Federal Register. The corporate notification letters issued after the effective date will provide facility licensees with at least 90 days notice that the examinations will be administered in accordance with the revised procedures

  1. A review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    In the frame of Italian nuclear program, this report deals with the current activities related to PEC reactor delay in construction and start-up, activities within the joint venture between Novatome, France and NIRA, Italy related to components for Super Phenix reactor, participation of NIRA in the Super Phenix studies covering technology of reactor components, reactor core, fuel, safety, fuel cycle technical and economical aspects, codes and standards

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  3. Safety analysis for non-power reactors

    International Nuclear Information System (INIS)

    Non-power reactors have been operating in Canada since 1945, with NRU (National Research Universal, 1957) being the oldest operating non-power reactor. Presently, there are five generic 'types' of non-power reactors: NRU, ZED-2, SLOWPOKE, MNR and MAPLE, the latter undergoing commissioning as the MDS Medical Isotope Reactor. These reactors range in thermal power from 200 Watts to more than 100 MW. Other non-power reactors are likely to be built for new applications and to replace older reactors. The uniqueness of each reactor, the wide range of power levels and the evolution of safety philosophy over time have lead to non-uniform practices for safety analysis. This non-uniformity may be a problem for the preparation by the licensee and review by the regulator of the safety analysis report required for licensing of the reactor facility. Clearly, there is no universally applicable practice, while at the same time, expectations for safety analyses have evolved in order to demonstrate higher levels of overall safety. This paper examines a new 'graded approach' to preparing the safety analysis report for reactors of diverse features but with a common standard of safety. It discusses necessary content, methods and the training and qualification of the safety analyst. (author)

  4. Nuclear research reactors

    International Nuclear Information System (INIS)

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)

  5. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    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

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  7. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  8. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  9. Status of conversion of NE standards to national consensus standards

    International Nuclear Information System (INIS)

    One major goal of the Nuclear Standards Program is to convert existing NE standards into national consensus standards (where possible). This means that an NE standard in the same subject area using the national consensus process. This report is a summary of the activities that have evolved to effect conversion of NE standards to national consensus standards, and the status of current conversion activities. In some cases, all requirements in an NE standard will not be incorporated into the published national consensus standard because these requirements may be considered too restrictive or too specific for broader application by the nuclear industry. If these requirements are considered necessary for nuclear reactor program applications, the program standard will be revised and issued as a supplement to the national consensus standard. The supplemental program standard will contain only those necessary requirements not reflected by the national consensus standard. Therefore, while complete conversion of program standards may not always be realized, the standards policy has been fully supported in attempting to make maximum use of the national consensus standard. 1 tab

  10. TRIGA Reactor Power Upgrading Analysis

    International Nuclear Information System (INIS)

    Reactor physics safety analysis supporting the power upgrading from 1MW to 2MW of a typical TRIGA Mark II reactor is presented for steady state and pulse operation. The analysis is performed for mixed core configuration consisting of two types of fuel elements: standard 8,5% or 12% stainless-steel clad fuel elements and LEU fuel elements (20% uranium concentration). The following reactor physics codes are applied: WIMS, TRIGAC, EXTERMINATOR, PULSTRI and TRISTAN. Results of the calculations are compared to experiments for steady state operation at 1 MW. The analysis shows that besides technical modifications of the core (installation of an additional control rod) also some strict administrative limitations have to be imposed on operational parameters (excess reactivity, pulse reactivity, core composition) to assure safe operation within design limits. (author)

  11. Business Opportunities for Small Reactors

    International Nuclear Information System (INIS)

    This report assesses the market potential and identifies a number of potential paths for developing the small nuclear reactor business. There are several potential opportunities identified and evaluated. Selecting a specific approach for the business development requires additional information related to a specific market and sources of capital to support the investment. If and how a market for small nuclear plants may develop is difficult to predict because of the complexity of the economic and institutional factors that will influence such development. Key factors are; economics, safety, proliferation resistance and investment risk. The economic and political interest of any of the identified markets is also dependent on successful demonstration of the safety and reliability of small nuclear reactor. Obtaining a US-NRC Standard Design approval would be an important development step toward establishing a market for small reactors. (authors)

  12. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  13. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  14. Assessment of thermal aging embrittlement of cast austenitic stainless steel components in the Babcock and Wilcox -designed PWR reactor internals

    International Nuclear Information System (INIS)

    The currently operating Babcock and Wilcox (BW) designed pressurized water reactors (PWRs) were constructed during the late sixties and seventies. Some of the reactor internals components were fabricated from cast austenitic stainless steel (CASS). The selection of CASS for the internals components was made to expedite the construction schedule by reducing machining and allowing production in large quantities. Since then, test data have shown that some CASS materials are susceptible to thermal aging embrittlement at PWR operating temperatures and its effect on functionality is of concern. Recently, the US nuclear power industry has developed inspection and evaluation guidelines (MRP-227, Rev.0) for managing aging degradation in PWR reactor internals for both the current and extended license periods. The MRP-227, Rev.0 guidelines recommend additional inspections for certain internals components including CASS components in BW PWRs due to thermal aging embrittlement concerns. The thermal aging embrittlement susceptibility for CASS can be assessed by the casting method and ferrite content if sufficient information in the original fabrication records is available. AREVA NP has performed a fabrication records search to identify several CASS components in the BW PWR internals and reviewed the archived fabrication records. A database has been assembled as a result of this records search. Based on the fabrication records, the ferrite content is determined using Hull's equivalent factors. Grade CF8 castings (without molybdenum) have been found to not be susceptible to thermal aging embrittlement. However, thermal aging embrittlement is a potential concern for Grade CF3M castings (containing 2 to 3% molybdenum). As a result of this assessment, several CASS components in the BW PWRs are concluded to not be susceptible to thermal aging embrittlement. The findings provide the basis for the removal of these CASS components from the additional inspection requirements in MRP-227

  15. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

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

  16. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-βN tokamak reactors. (author)

  17. TRIGA mark-II,III reactor safety re-evaluation

    International Nuclear Information System (INIS)

    For two years of 1990 and 1991, the safety of TRIGA Mk-II and III reactor has been re-evaluated. For this, domestic rules on research reactors has been reviewed, and as it was judged that standards on research reactors in USA is applicable to our ones it was evaluated whether TRIGA Mk-II and III reactors satisfy these standards. The site parameters and the environmental impacts during normal operation and hypothetical accident conditions have been analysed, and those parts for reactor facility and structure have been rewritten to fit SAR standard format based on the review of old SAR and maintenance manuals reflecting changes after the construction. Based on this re-evaluation, SAR, Technical Specifications, Radiation Emergency Plan, Environment Report, various procedures,etc. will be amended by the reactor management project. (Author)

  18. Economic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U3O8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented

  19. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.)

  20. ['Gold standard', not 'golden standard'

    NARCIS (Netherlands)

    Claassen, J.A.H.R.

    2005-01-01

    In medical literature, both 'gold standard' and 'golden standard' are employed to describe a reference test used for comparison with a novel method. The term 'gold standard' in its current sense in medical research was coined by Rudd in 1979, in reference to the monetary gold standard. In the same w

  1. Developments in the regulation of research reactors

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has data on over 670 research reactors in the world. Fewer than half of them are operational and a significant number are in a shutdown but not decommissioned state. The International Nuclear Safety Advisory Group (INSAG) has expressed concerns about the safety of many research reactors and this has resulted in a process to draw up an international Code of Conduct on the Safety of Research Reactors. The IAEA is also reviewing its safety standards applying to research reactors. On the home front, regulation of the construction of the Replacement Research Reactor continues. During the construction phase, regulation has centred around the consideration of Requests for Approval (RFA) for the manufacture and installation of systems, structures and components important for safety. Quality control of construction of systems, structures and components is the central issue. The process for regulation of commissioning is under consideration

  2. Decommissioning standards

    International Nuclear Information System (INIS)

    EPA has agreed to establish a series of environmental standards for the safe disposal of radioactive waste through participation in the Interagency Review Group on Nuclear Waste Management (IRG). One of the standards required under the IRG is the standard for decommissioning of radioactive contaminated sites, facilities, and materials. This standard is to be proposed by December 1980 and promulgated by December 1981. Several considerations are important in establishing these standards. This study includes discussions of some of these considerations and attempts to evaluate their relative importance. Items covered include: the form of the standards, timing for decommissioning, occupational radiation protection, costs and financial provisions. 4 refs

  3. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  4. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  5. RB reactor noise analysis

    International Nuclear Information System (INIS)

    Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed

  6. Basis for Interim Operation for the K-Reactor in Cold Standby

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, B.

    1998-10-19

    The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993).

  7. Basis for Interim Operation for the K-Reactor in Cold Standby

    International Nuclear Information System (INIS)

    The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993)

  8. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

  9. Accounting standards

    NARCIS (Netherlands)

    B. Stellinga

    2014-01-01

    The European and global regulation of accounting standards have witnessed remarkable changes over the past twenty years. In the early 1990s, EU accounting practices were fragmented along national lines and US accounting standards were the de facto global standards. Since 2005, all EU listed companie

  10. International Standards.

    Science.gov (United States)

    Havard-Williams, Peter

    1982-01-01

    Discussion of standardization on an international scale for resource sharing--cooperation, coordination, interlibrary loans, cooperative acquisition and cataloging--focuses on a definition of standards; the development of standards for cataloging; public, school, and university libraries; and library education. A 60-item bibliography is included.…

  11. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  12. Commission of the European Communities review of fast reactor activities, March 1981

    International Nuclear Information System (INIS)

    The Commission of the European Communities continued its activities in the field of fast reactors development essentially in the frame of the Fast Reactor Coordinating Committee (FRCC) and by execution of a Reactor Programme at its Joint Research Center (JRC). The study was concerned with introducing fast reactors into European Community, elaboration of preliminary safety criteria and guidelines for typical fast reactor accidents; codes and standards; LMFBR safety, fuel, fuel cycle safety

  13. Safety of Ghana Research Reactor (GHARR-1)

    International Nuclear Information System (INIS)

    The Ghana Research Reactor, GHARR-1 is a low power research rector with maximum thermal power lever of 30kW. The reactor is inherently safe and uses highly enriched uranium (HEU) as fuel, light water as moderator and beryllium as a reflector. The construction, commissioning and operation of this reactor have been subjected to the system of authorization and inspection developed by the Regulatory Authority, the Radiation Protection Board (RPB) with the assistance of the International Atomic Energy Agency. The reactor has been regulated by the preparation of an Interim Safety Analysis Report (SAR) based upon International Atomic Energy Agency standards. An International Safety Assessment peer review and safe inspections have confirmed a high level of operational safety of the reactor since it started operation in 1994. Since its operation there has been no significant reported incident/accidents. Several studies have validated the inherent safety of the reactor. The reactor has been used for neutron activation analysis of various samples, research and teaching. About 1000 samples are analysed annually. The final Safety Analysis Report (SAR) was submitted (after five years of extensive research on the operational reactor) to the Regulatory Authority for review in June 2000. (author)

  14. Reactor Physics Training

    International Nuclear Information System (INIS)

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

  15. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

  16. Research Reactors: Decommissioning of a Small Reactor (BR3 Reactor, Belgium). Appendix III

    International Nuclear Information System (INIS)

    Research reactors are nuclear reactors that serve primarily as source of neutrons. They are less complex than power reactors and operate at lower temperatures. Research reactors need far less fuel, and far less fission products build up as the fuel is used. On the other hand, their fuel requires more highly enriched uranium, typically up to 20% 235U. More than 650 research reactors worldwide have been built or are under construction or in a planning phase; of which more than 350 have been shut down and partly or wholly decommissioned. Experience has shown that decommissioning can be undertaken in line with safety standards aimed at protecting human beings or the environment from harm, provided that decommissioning activities are undertaken in accordance with a properly formulated plan. The potential or actual radiological hazards associated with reactors may require the application of special techniques and procedures during decommissioning. The decommissioning of the BR3 reactor in Mol, Belgium, Belgian nuclear research centre SCK•CEN, provides an example of current good practice in decommissioning research reactors.13 Since 1991, the organization’s statutory mission gives priority to research on problems of societal concern such as the safety of nuclear installations, radiation protection, safe treatment and disposal of radioactive waste, fighting against uncontrolled proliferation of fissile materials, and education and training. BR3 was the first European pressurized water reactor (PWR) power plant and was put into service in 1962. It was in that industrial context that the BR3 has played its role as a demonstration unit for the development and improvement of decommissioning related techniques. While the BR3 power level was low (40 MW(th), 10.5 MW(e) net), it contains all the features of commercial PWR power plants. The reactor was used at the beginning of its lifetime as a training facility for future nuclear power plant operators. Later, it was also used

  17. Reactor calculations for improving utilization of TRIGA reactor

    International Nuclear Information System (INIS)

    A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

  18. Safeguarding research reactors

    International Nuclear Information System (INIS)

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  19. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  20. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  1. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  2. SP-100 Reactor Subsystem Development

    Science.gov (United States)

    Demuth, Scott F.

    1994-07-01

    The SP-100 reactor subsystem consists of the pressure vessel, vessel internals, and fuel elements. Type A (standard) Nb-1Zr and rhenium materials development efforts related to fabrication of the vessel, vessel internals, and fuel cladding/liner have been completed. Type A and Type C (PWC-11) Nb-1Zr loop fabrication has been successfully demonstrated by prototypic testing with flowing lithium at 1350 K for 1500 hr. Development of UN fuel has been completed, and the performance validated by irradiation testing to the full life (7 yr. full power) burnup of 6 atom %. Neutronic and hydraulic core performance have been validated by engineering mockup critical experiments in the Zero Power Physics Reactor at Argonne National Laboratory, and detailed core hydraulic flow testing with water. Essentially all feasibility issues have been settled for the full life SP-100 reactor subsystem. Remaining SP-100 reactor subsystem development efforts are focused on further reducing mass by the use of Type C (PWC-11) Nb-1Zr rather than Type A, and demonstrating fuel life for beyond full life to perhaps 9 atom % burnup.

  3. Communications standards

    CERN Document Server

    Stokes, A V

    1986-01-01

    Communications Standards deals with the standardization of computer communication networks. This book examines the types of local area networks (LANs) that have been developed and looks at some of the relevant protocols in more detail. The work of Project 802 is briefly discussed, along with a protocol which has developed from one of the LAN standards and is now a de facto standard in one particular area, namely the Manufacturing Automation Protocol (MAP). Factors that affect the usage of networks, such as network management and security, are also considered. This book is divided into three se

  4. Reactor Physics Programme

    International Nuclear Information System (INIS)

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

  5. Review of ANS-19 standards and activities

    International Nuclear Information System (INIS)

    The standards subcommittee ANS-19, 'Physics of Reactor Design' sponsored by the ANS Reactor Physics Division, has been in existence since 1972. Five standards have been produced, two data review groups have been established, and a working group has reviewed reactor physics terms for inclusion in the N 1.1 American National Standard Glossary of Terms in Nuclear Science and Technology. Much to the disappointment of those of us involved in their development, these standards have for the most part been uniformly ignored by all segments of the nuclear industry. Nevertheless, there is still great interest in revising the standards to make them more useful. It is hoped that the changes now under consideration as well as possible new standards areas of physics inputs to safety and transient calculations and operational power reactor measurements for license compliance will help to bring this about. There is a general feeling that within the large and diverse nuclear industry there is real benefit to be gained both technically and in public perception by the introduction of increasing standardization of good engineering practices. Such standardization to have maximum benefit must come from the technical experts within the industry, rather than being imposed by the regulators. (orig.)

  6. Risk management activities at the DOE Class A reactor facilities

    International Nuclear Information System (INIS)

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented

  7. N Reactor Lessons Learned workshop

    International Nuclear Information System (INIS)

    This report describes a workshop designed to introduce participants to a process, or model, for adapting LWR Safety Standards and Analysis Methods for use on rector designs significantly different than LWR. The focus of the workshop is on the ''Lessons Learned'' from the multi-year experience in the operation of N Reactor and the efforts to adapt the safety standards developed for commercial light water reactors to a graphite moderated, water cooled, channel type reactor. It must be recognized that the objective of the workshop is to introduce the participants to the operation of a non-LWR in a LWR regulatory world. The total scope of this topic would take weeks to provide a through overview. The objective of this workshop is to provide an introduction and hopefully establish a means to develop a longer term dialogue for technical exchange. This report provides outline of the workshop, a proposed schedule of the workshop, and a description of the tasks will be required to achieve successful completion of the project

  8. Inconsistencies of neutron flux parameters for k(0) standardization in neutron activation analysis determined with the use of Au+Zr and Au+Mo+Cr monitor sets at the LVR-15 reactor in Rez

    Czech Academy of Sciences Publication Activity Database

    Kubešová, Marie; Kučera, Jan

    2012-01-01

    Roč. 293, č. 2 (2012), s. 665-674. ISSN 0236-5731 R&D Projects: GA ČR GA202/09/0363 Institutional support: RVO:61389005 Keywords : Neutron activation analysis * K(0) standardization * Neutron flux parameters Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.467, year: 2012

  9. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  10. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  11. Process heat reactors

    International Nuclear Information System (INIS)

    The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre

  12. Reactor System Design

    International Nuclear Information System (INIS)

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  13. Design guide for category II reactors light and heavy water cooled reactors

    International Nuclear Information System (INIS)

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems

  14. Design guide for Category III reactors: pool type reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

  15. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  16. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  17. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2016-01-01

    International e-Customs is going through a standardization process. Driven by the need to increase control in the trade process to address security challenges stemming from threats of terrorists, diseases, and counterfeit products, and to lower the administrative burdens on traders to stay...... competitive, national customs and regional economic organizations are seeking to establish a standardized solution for digital reporting of customs data. However, standardization has proven hard to achieve in the socio-technical e-Customs solution. In this chapter, the authors identify and describe what has...... to be harmonized in order for a global company to perceive e-Customs as standardized. In doing so, they contribute an explanation of the challenges associated with using a standardization mechanism for harmonizing socio-technical information systems....

  18. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2014-01-01

    International e-Customs is going through a standardization process. Driven by the need to increase control in the trade process to address security challenges stemming from threats of terrorists, diseases, and counterfeit products, and to lower the administrative burdens on traders to stay...... competitive, national customs and regional economic organizations are seeking to establish a standardized solution for digital reporting of customs data. However, standardization has proven hard to achieve in the socio-technical e-Customs solution. In this chapter, the authors identify and describe what has...... to be harmonized in order for a global company to perceive e-Customs as standardized. In doing so, they contribute an explanation of the challenges associated with using a standardization mechanism for harmonizing socio-technical information systems....

  19. Training Standardization

    International Nuclear Information System (INIS)

    The article describes the benefits of and required process and recommendations for implementing the standardization of training in the nuclear power industry in the United States and abroad. Current Information and Communication Technologies (ICT) enable training standardization in the nuclear power industry. The delivery of training through the Internet, Intranet and video over IP will facilitate this standardization and bring multiple benefits to the nuclear power industry worldwide. As the amount of available qualified and experienced professionals decreases because of retirements and fewer nuclear engineering institutions, standardized training will help increase the number of available professionals in the industry. Technology will make it possible to use the experience of retired professionals who may be interested in working part-time from a remote location. Well-planned standardized training will prevent a fragmented approach among utilities, and it will save the industry considerable resources in the long run. It will also ensure cost-effective and safe nuclear power plant operation

  20. The Chernobylsk reactor accident

    International Nuclear Information System (INIS)

    The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)

  1. Zero energy reactor 'RB'

    International Nuclear Information System (INIS)

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  2. Thermal stratification of sodium in the BN 600 reactor

    International Nuclear Information System (INIS)

    The signs of thermal stratification of sodium in the BN 600 reactor upper plenum revealed by the analysis of standard temperature sensors' readings are defined. The initial conditions for existence of different temperature sodium layers are given. Two approaches for realizing on a computer of equations describing sodium motion in the upper plenum of the reactor are presented. (author)

  3. High solids fermentation reactor

    Science.gov (United States)

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  4. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  5. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  6. Reactor power measuring device

    International Nuclear Information System (INIS)

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  7. Design guide for category VI reactors: air-cooled graphite reactors

    International Nuclear Information System (INIS)

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned air-cooled graphite reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  8. Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

    International Nuclear Information System (INIS)

    RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

  9. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  10. Designing a reactor for the next generation

    International Nuclear Information System (INIS)

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept being developed under a cooperative program involving the U.S. Government, the utilities and the nuclear industry. The design utilizes basic High-Temperature Gas-Cooled Reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The specific size and configuration is selected to utilize the inherent materials characteristics associated with these standard features to develop a passively safe design which provides a higher margin of safety and investment protection than current generation reactors. The MHTGR which is now into the preliminary design phase meets regulatory and new, more challenging user requirements. This paper outlines the requirements, describes the design and the current status, shows how the design meets the requirements, and looks at potential future deployment from the vendor's perspective

  11. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  12. Reactor Physics Modeling Of Spent Nuclear Research Reactor Fuel For SNM Attribution And Nuclear Forensics

    International Nuclear Information System (INIS)

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  13. Management system requirements for small reactors

    International Nuclear Information System (INIS)

    This abstract identifies the management system requirements for the life cycle of small reactors from initial conception through completion of decommissioning. For small reactors, the requirements for management systems remain the same as those for 'large' reactors regardless of the licensee' business model and objectives. The CSA N-Series of standards provides an interlinked set of requirements for the management of nuclear facilities. CSA N286 provides overall direction to management to develop and implement sound management practices and controls, while other CSA nuclear standards provide technical requirements and guidance that support the management system. CSA N286 is based on a set of principles. The principles are then supported by generic requirements that are applicable to the life cycle of nuclear facilities. CNSC regulatory documents provide further technical requirements and guidance. (author)

  14. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  15. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  16. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  17. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  18. EOS standards

    Energy Technology Data Exchange (ETDEWEB)

    Greeff, Carl W [Los Alamos National Laboratory

    2011-01-12

    An approach to creating accurate EOS for pressure standards is described. Applications to Cu, Au, and Ta are shown. Extension of the method to high compressions using DFT is illustrated. Comparisons with modern functionals show promise.

  19. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  20. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  1. High temperature reactors

    International Nuclear Information System (INIS)

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

  2. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  3. Quality assurance plan, Westinghouse Water Reactor Divisions

    Energy Technology Data Exchange (ETDEWEB)

    1976-03-01

    The Quality Assurance Program used by Westinghouse Nuclear Energy Systems Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements.

  4. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

    The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel

  5. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  6. 40 CFR 63.1407 - Non-reactor batch process vent provisions.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year...

  7. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  8. System 80+{trademark} Standard Design: CESSAR design certification. Volume 4: Amendment I

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-21

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 4, provides a description of the reactor, reactor internals, fuel assemblies, and associated design requirements.

  9. Germany: the government has approved of the extension of the service life of nuclear reactors

    International Nuclear Information System (INIS)

    The German coalition government has decided on 28 september 2009 to expend the service life of the 17 operating nuclear reactors through the approval of the plan 'Energy 2050'. The average extension will be of 12 years. This decision will postpone the shutdown of the last reactor till 2040. The reciprocal arrangement implies that the nuclear plant operators (EON, RWE, Wattenfall and EnBW) will have to pay an annual tax of 2.3*109 euros till 2016. The operators will also have to finance a capital for the development of renewable energies: 200*106 euros in 2011 and in 2012 then 300*106 euros per year between 2013 and 2016 and eventually to pay a tax of 9 euros per MW of nuclear power produced. (A.C.)

  10. Water cooled FBNR nuclear reactor

    International Nuclear Information System (INIS)

    Full text: The world with its increasing population and the desire for a more equitable and higher standard of living, is in the search for energy that is abundant and does not contribute to the problem of global warming. The answer to this is a new paradigm in nuclear energy; i.e., through the innovative nuclear reactors that meet the IAEA's INPRO philosophies and criteria that will guarantee the generation of safe and clean energy. The emerging countries to nuclear energy that are not in hurry for energy and look into the future are looking into the participation in the development of such innovative nuclear reactors. They can start developing the non-nuclear components of such reactors in parallel with creating the nuclear infra-structures according to the guidelines of the IAEA suggested in its milestones document. In this way, they can benefit from numerous advantages that the development of a high technology can bring to their countries be it scientific, technological, economic or political. A solution to the present world economic crisis is investing in such projects that contribute to the real economy rather than speculative economy. This will help both local and world economy. One such innovative nuclear reactor is the FBNR that is being developed with the support of the IAEA in its program of Small Reactors Without On-site Refuelling. It is a small (70 MWe) reactor with simple design based on the proven PWR technology (www.sefidvash.net/fbnr). The simplicity in design and the world wide existence of water reactor technology, makes it a near term project compared to other future reactors. Small reactors are most adequate for both the developing and developed countries. They require low capital investment, and can be deployed gradually as energy demand calls for. The generation of energy at the local of consumption avoids high cost of energy transmission. The paradigm of economy of scale does not apply to the FBNR as it is a small reactor by its nature. The

  11. IAEA Guidelines for New Research Reactor Projects

    International Nuclear Information System (INIS)

    'Technical Requirements in the Bidding Process for a New Research Reactor' (Nuclear Energy Series Report NP-T-5.6) was developed and is currently under publication process. This document is meant to bridge the gap between the feasibility study for a new research reactor and the call for bids. The publication also addresses the preparation phase of the bidding process and discusses criteria that may be used in the evaluation of the bids. The guidance applies to all reactor types and technologies and it is not recommending a specific reactor type or technology or a specific design. However, it is assumed that the publication will be used by a Member State that has made a knowledgeable commitment to build a safe, sustainable, robust-design and easy-maintainable research reactor. While the guidance provided by this publication is intended mainly to be used by Member States building their first research reactor, it might be suitable for Member States building a subsequent one. The above mentioned publications should be used in conjunction with the IAEA publications on research reactor safety and utilization, in particular the Code of Conduct on the Safety of Research Reactors and the supporting IAEA Safety Standards. This paper presents an overview of the publications and discusses follow-up activities with Member States for their effective application

  12. Research reactors in Argentina

    International Nuclear Information System (INIS)

    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

  13. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Some aspects of reactor operation management are highlighted. The main mission of the operational staff at a testing reactor is to operate it safely and efficiently, to ensure proper conditions for different research programs implying the use of the reactor. For reaching this aim, there were settled down operating plans for every objective, and procedure and working instructions for staff training were established, both for the start-up and for the safe operation of the reactor. Damages during operation or special situations which can arise, at stop, start-up, maintenance procedures were thoroughly considered. While the technical skill is considered to be the most important quality of the staff, the organising capacity is a must in the operation of any nuclear facility. Staff training aims at gaining both theoretical and practical experience based on standards about staff quality at each work level. 'Plow' sheet has to be carefully done, setting clear the decision responsibility for each person so that everyone's own technical level to be coupled to the problems which implies his responsibility. Possible events which may arise in operation, e.g., criticality, irradiation, contamination, and which do not arise in other fields, have to be carefully studied. One stresses that the management based on technical and scientific arguments have to cover through technical, economical and nuclear safety requirements a series of interlinked subprograms. Every such subprograms is subject to some peculiar demands by the help of which the entire activity field is coordinated. Hence for any subprogram there are established the objectives to be achieved, the applicable regulations, well-defined responsibilities, training of the personnel involved, the material and documentation basis required and activity planning. The following up of positive or negative responses generated by experiments and the information synthesis close the management scope. Important management aspects

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

    International Nuclear Information System (INIS)

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

  15. Thai research reactor

    International Nuclear Information System (INIS)

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

  16. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  17. EURATOM's Programme of Participation in Power Reactor Construction

    International Nuclear Information System (INIS)

    One of the means used by the Commission of EURATOM to promote the development of a European nuclear industry is a programme of ''Community participation'', under which the Commission will participate in power reactor construction up to a total expenditure of 32 million European Monetary Agreement units of account. The return for this will be the acquisition of information on the design, construction, start-up and operation of such reactors. So far, proposals from three companies have resulted in the signing of contracts. These companies are: (a) The Societa Elettronucleare Nazionale (SENN), which is constructing a station of 150 MW(e) net in Italy, equipped with a double-cycle boiling-water reactor; (b) The Societa Italiana Meridionale Energia Atomica (SIMEA), which has undertaken to construct a station f 200 MW(e) net in Italy, equipped with a natural uranium-graphite-CO2 reactor; (c) The Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA), which has undertaken to construct, on the French-Belgian border, a station which will be equipped with a pressurized-water reactor and whose output will reach, and probably exceed, 242 MW(e) net. Further, the Commission has been requested by the Rheinisch-Westfalisches Elektrizitatswerk - Bayernwerke (RWE-BW) group and the N.V. Samenwerkende Electriciteits-Productiebedrijve to take part in the construction o f two other power reactors - the first a 237 MW(e) double-cycle boiling-water reactor, and the second a 50 MW(e) single-cycle, natural-circulation boiling-water reactor. Community participation can take various forms, one of them being the sharing of any deficit that might result from the production of electricity by the stations during their first years of operation. The effect of EURATOM's participation has been to encourage the construction of some of these nuclear power stations. Moreover, it has resulted in the gathering of extremely useful information and w ill continue to do so in the years to come

  18. LASIP-III, a generalized processor for standard interface files. [For creating binary files from BCD input data and printing binary file data in BCD format (devised for fast reactor physics codes)

    Energy Technology Data Exchange (ETDEWEB)

    Bosler, G.E.; O' Dell, R.D.; Resnik, W.M.

    1976-03-01

    The LASIP-III code was developed for processing Version III standard interface data files which have been specified by the Committee on Computer Code Coordination. This processor performs two distinct tasks, namely, transforming free-field format, BCD data into well-defined binary files and providing for printing and punching data in the binary files. While LASIP-III is exported as a complete free-standing code package, techniques are described for easily separating the processor into two modules, viz., one for creating the binary files and one for printing the files. The two modules can be separated into free-standing codes or they can be incorporated into other codes. Also, the LASIP-III code can be easily expanded for processing additional files, and procedures are described for such an expansion. 2 figures, 8 tables.

  19. Review of neutronic assessments applied to small reactor core physics

    International Nuclear Information System (INIS)

    In its design division for material test reactors and research reactors, AREVA TA has to characterize these manufactured cores. This step is sequential with neutronics benchmarks associated with validation (standard Verification and Validation approach). The previous two points are embedded in core projects and can be run separately especially when experimental tests are foreseen for validation database enrichment. Methodological standard is given in order to match validation and benchmark process illustrated alongside with two specific items on critical research reactors (AZUR - JHR) and subcritical mock up (AZUR). (author)

  20. Projected Standard on neutron skyshine

    International Nuclear Information System (INIS)

    Current interest in neutron skyshine arises from the application of dry fuel handling and storage techniques at reactor sites, at the proposed monitored retrievable storage facility and at other facilities being considered as part of the civilian radioactive waste management programs. The chairman of Standards Subcommittee ANS-6, Radiation Protection and Shielding, has requested that a work group be formed to characterize the neutron skyshine problem and, if necessary, prepare a draft Standard. The work group is comprised of representatives of storage cask vendors, architect engineering firms, nuclear utilities, the academic community and staff members of national laboratories and government agencies. The purpose of this presentation summary is to describe the activities of the work group and the scope and contents of the projected Standard, ANS-6.6.2, ''Calculation and Measurement of Direct and Scattered Neutron Radiation from Nuclear Power Operations.'' The specific source under consideration by the work group is an array of dry fuel casks located at a reactor site. However, it is recognized that the scope of the standard should be broad enough to encompass other neutron sources. The Standard will define appropriate methodology for properly characterizing the neutron dose due to skyshine. This dose characterization is necessary, for example, in demonstrating compliance with pertinent regulatory criteria

  1. Determination of trace elements in standard reference materials by the ko-standardization method

    International Nuclear Information System (INIS)

    The ko-standardization method is suitable for routine multielement determinations by reactor neutron activation analysis (NAA). Investigation of NIST standard reference materials SRM 1571 Orchard Leaves, SRM 1572 Citrus leaves, and SRM 1573 Tomato Leaves showed the systematic error of 12 certified elements determined to be less than 8%. Thirty-four elements were determined in NIST proposed SRM 1515 Apple Leaves

  2. Dosimetry standards

    International Nuclear Information System (INIS)

    The following leaflets are contained in this folder concerning the National Physical Laboratory's measurement services available in relation to dosimetry standards: Primary standards of X-ray exposure and X-ray irradiation facilities, X-ray dosimetry at therapy levels, Protection-level X-ray calibrations, Therapy-level gamma-ray facility, Fricke dosemeter reference service, Low-dose-rate gamma-ray facility, Penetrameter and kV meter calibration, Measurement services for radiation processing, Dichromate dosemeter reference service, Electron linear accelerator. (U.K.)

  3. Frequency standards

    CERN Document Server

    Riehle, Fritz

    2006-01-01

    Of all measurement units, frequency is the one that may be determined with the highest degree of accuracy. It equally allows precise measurements of other physical and technical quantities, whenever they can be measured in terms of frequency.This volume covers the central methods and techniques relevant for frequency standards developed in physics, electronics, quantum electronics, and statistics. After a review of the basic principles, the book looks at the realisation of commonly used components. It then continues with the description and characterisation of important frequency standards

  4. Argentine nuclear energy standardization activities

    International Nuclear Information System (INIS)

    The International Organization for Standardization (ISO) has more than 200 Technical Committees that develop technical standards. During April 2004 took place in Buenos Aires the 14th Plenary of the ISO/TC 85 Nuclear Energy Committee. During this Plenary issues as Nuclear Terminology, Radiation Protection, Nuclear Fuels, Nuclear Reactors and Irradiation Dosimetry was dealt with. 105 International delegates and 45 National delegates (belonging to CNEA, ARN, NASA, INVAP, CONUAR, IONICS and other organizations) attended the meetings. During this meeting ISO/TC 85 changed its scope; the new scope of the Committee is 'Standardization in the fields of peaceful applications of nuclear energy and of the protection of individuals against all sources of ionizing radiations'. This work summarizes the most important advances and resolutions about the development of standards taken during this meeting as well as the main conclusions. (author)

  5. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  6. Proceedings of the specialists' meeting on reactor group constants

    International Nuclear Information System (INIS)

    This report is the Proceedings of the Specialists' Meeting on Reactor Group Constants. The meeting was held on February 22-23, 2001 at Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of 59 specialists. The evaluation work for JENDL-3.3 is going on for the publication in a short time. The processing JENDL-3.3 file to make reactor group constants is needed when it is used in application fields. In the meeting, the present status of the reactor group constants was reviewed and the issues relating to them were discussed in such fields as thermal reactor, criticality safety, fast reactor, high energy region, burn-up calculation and radiation shielding. At the final session in the meeting, standardization of reactor group constants was discussed and the need of the reference group constants was confirmed by the participants. The 11 of the presented papers are indexed individually. (J.P.N.)

  7. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACRTM-1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  8. Design and fabrication of HTTR reactor pressure vessel

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is under construction at the Oarai Research Establishment, JAERI and planned to be critical at the end of 1997. The HTTR is a High Temperature Gas-cooled Reactors (HTGRs) with thermal output of 30MW, inlet coolant temperature of 395degC, and outlet coolant temperature of 850degC at rated operation and 950degC at high temperature test operation. 2.25Cr-1Mo steel is chosen for the reactor pressure vessel of the HTTR because its temperature reaches about 400degC at normal operation. 2.25Cr-1Mo steel has higher creep rupture strength than Mn-Mo steel used for the reactor pressure vessels of Light Water Reactors (LWRs). For the components of the HTTR reactor pressure vessel subjected to low temperatures where creep deformation is negligible, a design guideline based on Japanese structural design standard for LWRs 'Technical standards for LWR power plant components-Ministry of International Trade and Industry Standard No.501' is utilized. On the other hand, design of the components for high temperature application, where creep behavior dominates, is conducted under newly determined high temperature structural design guideline and design material data. The fabrication of the HTTR reactor pressure vessel took about 23 months. It was installed in a reactor containment vessel in August, 1994. After core components had been installed in the reactor pressure vessel, pressure test of the primary and secondary cooling system including the reactor pressure vessel was performed and successfully ended in March, 1996. This paper reports issues of the HTTR reactor pressure vessel such as structure, material, stress analysis, fabrication, examination and testing. (author)

  9. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book Canada Enters the Nuclear Age. The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

  10. Evaluation of research reactors

    International Nuclear Information System (INIS)

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  11. Multipurpose research reactors

    International Nuclear Information System (INIS)

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  12. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  13. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  14. Laser standards

    International Nuclear Information System (INIS)

    The history of the development of international, American and British standards for the use of lasers is briefly discussed. Other topics briefly discussed include the biological effects of laser radiation, hazard classification systems for laser systems, maximum permissible exposures and radiation protection measures in practical considerations. (UK)

  15. Reactor physics experiments with thorium based clusters in AHWR - critical facility

    International Nuclear Information System (INIS)

    AHWR - Critical Facility (AHWR - CF) is a 'zero power' reactor designed to carry out various reactor physics experiments for validation of AHWR design. A number of experiments have been carried out in standard and extended reference core of the reactor. In this paper, results of experiments with different Thorium based experimental clusters are presented. These experiments provided valuable data for validation of reactor physics design methodologies. (author)

  16. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  17. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  18. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  19. The replacement research reactor

    International Nuclear Information System (INIS)

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  20. PFBR reactor protection

    International Nuclear Information System (INIS)

    Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy

  1. Reactor BR2

    International Nuclear Information System (INIS)

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  2. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  3. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  4. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  5. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  6. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  7. New reactor concepts

    International Nuclear Information System (INIS)

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  8. Reactor construction steels

    International Nuclear Information System (INIS)

    The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)

  9. Commercialization of fast reactors

    International Nuclear Information System (INIS)

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  10. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)

  11. INVAP's Research Reactor Designs

    International Nuclear Information System (INIS)

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety

  12. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention provides a control device which can conduct scram and avoid lowering of the power of a nuclear power plant upon occurrence of earthquakes. Namely, the device of the present invention comprises, in addition to an existent power control device, (1) an earthquake detector for detecting occurrence and annihilation of earthquakes and (2) a reactor control device for outputting control rod operation signals and reactor core flow rate control signals depending on the earthquake detection signals from the detector, and reactor and plant information. With such a constitution, although the reactor is vibrated by earthquakes, the detector detects slight oscillations of the reactor by initial fine vibration waves as premonitory symptoms of serious earthquakes. The earthquake occurrence signals are outputted to the reactor control device. The reactor control device, receiving the signals, changes the position of control rods by way of control rod driving mechanisms to make the axial power distribution in the reactor core to a top peak type. As a result, even if the void amount in the reactor core is reduced by the subsequent actual earthquakes, since the void amount is moved, effects on the increase of neutron fluxes by the actual earthquakes is small. (I.S.)

  13. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  14. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  15. Pebble bed modular reactor (PBMR)

    International Nuclear Information System (INIS)

    accidents without exceeding the failure temperature of the coated particles, which is key to the normal safety characteristics of all HTGRs. Originally, the focus of the modular HTGR was on the steam cycle and included designs by Germany, the Russian Federation and the USA. These designs all incorporated the TRISO ceramic coated fuel particles and utilized steel vessels to house the primary system. Design of the present direct cycle gas turbine modular plant such as the PBMR began in the early 1990s. This plant incorporates the basic safety attributes of the modular pebble bed reactor with the direct improvement of not being tied to the complexities and low efficiencies associated with the steam cycle. Also attendant with the modular direct cycle PBMR is a high degree of standardization for this relatively small, simplified design. This approach allows the benefits of plant modularization and shop fabrication with corresponding improvements in quality control, reduction of construction schedules, and optimization of manufacturing procedures and processes; all resulting in improvements in schedule and capital costs. A listing is given of the more prominent test facilities and R and D performed to support the PBMR

  16. New research reactor for Australia

    International Nuclear Information System (INIS)

    HIFAR, Australia's major research reactor was commissioned in 1958 to test materials for an envisaged indigenous nuclear power industry. HIFAR is a Dido type reactor which is operated at 10 MW. With the decision in the early 1970's not to proceed to nuclear power, HIFAR was adapted to other uses and has served Australia well as a base for national nuclear competence; as a national facility for neutron scattering/beam research; as a source of radioisotopes for medical diagnosis and treatment; and as a source of export revenue from the neutron transmutation doping of silicon for the semiconductor industry. However, all of HIFAR's capabilities are becoming less than optimum by world and regional standards. Neutron beam facilities have been overtaken on the world scene by research reactors with increased neutron fluxes, cold sources, and improved beams and neutron guides. Radioisotope production capabilities, while adequate to meet Australia's needs, cannot be easily expanded to tap the growing world market in radiopharmaceuticals. Similarly, neutron transmutation doped silicon production, and export income from it, is limited at a time when the world market for this material is expanding. ANSTO has therefore embarked on a program to replace HIFAR with a new multi-purpose national facility for nuclear research and technology in the form of a reactor: a) for neutron beam research, - with a peak thermal flux of the order of three times higher than that from HIFAR, - with a cold neutron source, guides and beam hall, b) that has radioisotope production facilities that are as good as, or better than, those in HIFAR, c) that maximizes the potential for commercial irradiations to offset facility operating costs, d) that maximizes flexibility to accommodate variations in user requirements during the life of the facility. ANSTO's case for the new research reactor received significant support earlier this month with the tabling in Parliament of a report by the Australian Science

  17. Safety Management at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Adequate safety measures and precautions, which follow relevant safety standards and procedures, should be in place so that personnel safety is assured. Nevertheless, the public, visitor, contractor or anyone who wishes to enter or be in the reactor building should be well informed with the safety measures applied. Furthermore, these same elements of safety are also applied to other irradiation facilities within the premises of Nuclear Malaysia. This paper will describes and explains current safety management system being enforced especially in the TRIGA PUSPATI Reactor (RTP) namely radiation monitoring system, safety equipment, safe work instruction, and interconnected internal and external health, safety and security related departments. (author)

  18. PUSPATI Triga reactor fuel worth measurement

    International Nuclear Information System (INIS)

    The reactivity worth of fuel elements in the B, C, D, E and F rings in the PUSPATI TRIGA Reactor core with respect to water as well as that of dummy fuel element (graphite filled) in the G ring were measured. The reactivity worth of 8.5 w/o standard TRIGA fuel element with respect to the dummy element in the B to F rings were also determined. The measured results agreed with the typical values given by the reactor supplier, General Atomatic Company, to within eight percents. (author)

  19. Actinides recycling assessment in a thermal reactor

    International Nuclear Information System (INIS)

    Highlights: • Actinides recycling is assessed using BWR fuel assemblies. • Four fuel rods are substituted by minor actinides rods in a UO2 and in a MOX fuel assembly. • Performance of standard fuel assemblies and the ones with the substitution is compared. • Reduction of actinides is measured for the fuel assemblies containing minor actinides rods. • Thermal reactors can be used for actinides recycling. - Abstract: Actinides recycling have the potential to reduce the geological repository burden of the high-level radioactive waste that is produced in a nuclear power reactor. The core of a standard light water reactor is composed only by fuel assemblies and there are no specific positions to allocate any actinides blanket, in this assessment it is proposed to replace several fuel rods by actinides blankets inside some of the reactor core fuel assemblies. In the first part of this study, a single uranium standard fuel assembly is modeled and the amount of actinides generated during irradiation is quantified for use it as reference. Later, in the same fuel assembly four rods containing 6 w/o of minor actinides and using depleted uranium as matrix were replaced and depletion was simulated to obtain the net reduction of minor actinides. Other calculations were performed using MOX fuel lattices instead of uranium standard fuel to find out how much reduction is possible to obtain. Results show that a reduction of minor actinides is possible using thermal reactors and a higher reduction is obtained when the minor actinides are embedded in uranium fuel assemblies instead of MOX fuel assemblies

  20. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  1. Safety considerations for research reactors in extended shutdown

    International Nuclear Information System (INIS)

    According to the IAEA Research Reactor Database, in the last 20 years, 367 research reactors have been shut down. Of these, 109 have undergone decommissioning and the rest are in extended shutdown with no clear definition about their future. Still other research reactors are infrequently operated with no meaningful utilization programme. These two situations present concerns related to safety such as loss of corporate memory, personnel qualification, maintenance of components and systems and preparation and maintenance of documentation. There are many reasons to shut down a reactor; these may include: - the need to carry out modifications in the reactor systems; - the need for refurbishment to extend the lifetime of the reactor; - the need to repair reactor structures, systems, or components; - the need to remedy technical problems; - regulatory or public concerns; - local conflicts or wars; - political convenience; - the lack of resources. While any one of these reasons may lead to shutdown of a reactor, each will present unique problems to the reactor management. The large variations from one research reactor to the next also will contribute to the uniqueness of the problems. Any option that the reactor management adopts will affect the future of the facility. Options may include dealing with the cause of the shutdown and returning to normal operation, extending the shutdown period waiting a future decision, or decommissioning. Such options are carefully and properly analysed to ensure that the solution selected is the best in terms of reactor type and size, period of shutdown and legal, economic and social considerations. This publication provides information in support of the IAEA safety standards for research reactors

  2. Safety evaluation report related to the preliminary design of the Standard Reference System, RESAR-414

    International Nuclear Information System (INIS)

    The safety evaluation for the Westinghouse Standard Reactor includes information on general reactor characteristics; design criteria for systems and components; reactor coolant system; engineered safety systems; instrumentation and controls; electric power systems; auxiliary systems; steam and power conversion system; radioactive waste management; radiation protection; conduct of operations; accident analyses; and quality assurance

  3. One piece reactor removal

    International Nuclear Information System (INIS)

    Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)

  4. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  5. The fusion reactor

    International Nuclear Information System (INIS)

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  6. Polymerization Reactor Engineering.

    Science.gov (United States)

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  7. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  8. Gas-cooled reactors

    International Nuclear Information System (INIS)

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  9. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  10. Light water type reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  11. Standards for maintenance documentation of COSMOS programs

    International Nuclear Information System (INIS)

    The Compatible Open Shop Modular Operating Scheme (COSMOS) is used for fast reactor neutronics and subassembly distortion calculations in the UK. It provides database and databank facilities for this purpose. A large number of applications programs use these facilities. This report sets down the standards agreed for use in the preparation of maintenance documents for these programs. These standards have been developed as the result of experience with earlier more complicated standards. Some of the terminology used in this report is particular to COSMOS, but the general features of the standard may be useful to those responsible for other computer programs. (author)

  12. Analysis of standard substance human hair

    International Nuclear Information System (INIS)

    The human hair samples as standard substances were analyzed by the neutron activation analysis (NAA) on the miniature neutron source reactor. 19 elements, i.e. Al, As, Ba, Br, Ca, Cl, Cr, Co, Cu, Fe, Hg, I, Mg, Mn, Na, S, Se, V and Zn, were measured. The average content, standard deviation, relative standard deviation and the detection limit under the present research conditions were given for each element, and the results showed that the measured values of the samples were in agreement with the recommended values, which indicated that NAA can be used to analyze standard substance human hair with a relatively high accuracy. (authors)

  13. Naval propulsion reactors

    International Nuclear Information System (INIS)

    This article deals with the design and exploitation of naval propulsion reactors, mainly of PWR-type. The other existing or conceivable types of reactors are also presented: 1 - specificities of nuclear propulsion (integration in the ship, marine environment, maneuverability, instantaneous availability, conditions of exploitation-isolation, nuclear safety, safety authority); 2 - PWR-type reactor (stable operation, mastered technology, general design, radiation protection); 3 - other reactor types; 4 - compact or integrated loops architecture; 5 - radiation protection; 6 - reactor core; 7 - reactivity control (core lifetime, control means and mechanisms); 8 - core cooling (natural circulation, forced circulation, primary flow-rate program); 9 - primary loop; 10 - pressurizer; 11 - steam generators and water-steam secondary loop; 12 - auxiliary and safety loops; 13 - control instrumentation; 14 - operation; 15 - nuclear wastes and dismantling. (J.S.)

  14. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  15. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 1014 n/cm2/sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  16. Reactor core monitoring method

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Michitsugu [Tokyo Electric Power Co., Inc. (Japan); Kanemoto, Shigeru; Enomoto, Mitsuhiro; Ebata, Shigeo

    1998-05-06

    The present invention provides a method of monitoring the state of coolant flow in a reactor of a BWR power plant. Namely, a plurality of local power region monitors (LPRM) are disposed to the inside of the reactor core for monitoring a power distribution. Signals of at least two optional LPRM detectors situated at positions different in axial or radial positions of the reactor core are obtained. General fluctuation components which nuclear hydrothermally fluctuate in overall reactor core are removed from the components of the signals. Then, correlational functions between these signals are determined. The state of coolant flow in the reactor is monitored based on the correlational function. When the axial flowing rate and radial flow interference are monitored, the accuracy upon monitoring axial and radial local behaviors of coolants can be improved by thus previously removing the general fluctuation components from signals of LPRM detectors and extracting local void information near to LPRM detectors at high accuracy. (I.S.)

  17. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  18. Mirror reactor surface study

    International Nuclear Information System (INIS)

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  19. FBR type reactor

    International Nuclear Information System (INIS)

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  20. TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  1. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  2. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  3. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  4. Iodine chemistry in a reactor regulation

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A. [Nuclear Regulatory Commission, Washington, DC (United States). Advisory Committee on Reactor Safeguards

    1996-12-01

    Radioactive iodine has always been an important consideration in the regulation of nuclear power reactors to assure the health and safety of the public. Regulators adopted conservatively bounding predictions of iodine behavior in the earliest days of the development of nuclear power because there was so little known about either accidents or the chemistry of iodine. Today there is a flood of new information and understanding of the chemistry of iodine under reactor accident conditions. This paper offers some thoughts on how the community of scientists engaged in the study of iodine chemistry can present the results of their work so that it is more immediately adopted by the regulator. It is suggested that the scientific community consider the concept of consensus standards so effectively used within the engineering community to define the status of the study of radioactive iodine chemistry for reactor safety. (author) 9 refs.

  5. Development of smart nuclear instrumentation for reactors

    International Nuclear Information System (INIS)

    Variety of nuclear instruments are required for different applications in reactors such as reactor start-up, reactor protection and regulating system, area monitoring, failed fuel detection, stack monitoring etc. Attempts are made to develop a standardized microcomputer based hardware for configuring different types of instruments. PC architecture is chosen due to easy availability of components/boards and software. These instruments have dual redundant Network Interface Cards for connecting to a Primary Radiation Data LAN which in turn can be connected to Plant Information Bus through Gateways. These SMART instruments can be tested/calibrated through specific commands from remote computers connected over the LAN. This paper describes the various issues involved and the design details. (author)

  6. Pakistan research reactor-1 and its upgradation

    International Nuclear Information System (INIS)

    In this article the author describes the procedure of renovation and upgradation of a swimming pool type Pakistan Research Reactor-1 (PARR-1) installed at PINSTECH. The reactor originally designed for a thermal power of 5 MW using highly enriched uranium as has been upgraded 10 MW with low enriched uranium as fuel. All the required safety precaution has been also modified with the new requirements. The cooling system of PARR-1 was modified to meet the requirements of upgraded power of 10 MW. In order to ensure safety for upgraded PARR-1 and to bring the reactor the current safety standards, some additional safety systems have been provided. An emergency core cooling system ECCS has been installed to remove core decay heat in case of loss of coolant accident (LOCA). (A.B.)

  7. Standard deviations

    CERN Document Server

    Smith, Gary

    2015-01-01

    Did you know that having a messy room will make you racist? Or that human beings possess the ability to postpone death until after important ceremonial occasions? Or that people live three to five years longer if they have positive initials, like ACE? All of these ‘facts' have been argued with a straight face by researchers and backed up with reams of data and convincing statistics.As Nobel Prize-winning economist Ronald Coase once cynically observed, ‘If you torture data long enough, it will confess.' Lying with statistics is a time-honoured con. In Standard Deviations, ec

  8. Multi-purpose reactor

    International Nuclear Information System (INIS)

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MWth, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co60) production capacity is 50000 Ci/yr, 200Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  9. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author)

  10. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  11. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  13. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  14. Handover standards.

    Science.gov (United States)

    2016-05-01

    An important part of discharge communication is the timely handover of information about diagnostic tests, as breakdown in this aspect of communication can contribute to unsafe patient care. NHS England has produced a set of standards to underpin the development of robust systems of care, policies and practice for the safe and high quality transfer of information about diagnostic tests and test results at discharge. The standards are governed by three overarching principles that have implications for nurses. They are that: ■ Clinicians who order tests are responsible for reviewing, acting on and communicating results and actions taken to GPs and patients, even if patients have been discharged. ■ Results received by GP practices should be reviewed and acted on by a responsible clinician even if they did not order the tests. ■ Reasonable adjustments should be made for people with learning disabilities and mental health problems and, where appropriate, families, carers, care co-ordinators and key workers should be invited to participate in handover processes and decisions about patients at discharge. PMID:27138516

  15. Development of advanced nuclear reactors in Russia

    International Nuclear Information System (INIS)

    Several advanced reactor designs have been so far developed in Russia. The AES-91 and AES-92 plants with the VVER-1000 reactors have been developed at the beginning of 1990. However, the former design has been built in China and the latest which is certified meeting European Utility Requirements is being built in India. Moreover, the model VVER-1500 reactor with 50-60 MWd/t burn-up and an enhanced safety was being developed by Gidropress about 2005, excepting to be completed in 2007. But, this schedule has slipped in favor of development of the AES-2006 power plant incorporating a third-generation standardized VVER-1200 reactor of 1170 MWe. This is an evolutionary development of the well-proven VVER-1000 reactor in the AES-92 plant, with longer life, greater power and efficiency and its lead units are being built at Novovoronezh II, to start operation in 2012-13. Based on Atomenergoproekt declaration, the AES-2006 conforms to both Russian standards and European Utility Requirements. The most important features of the AES-2006 design are mentioned as: a design based on the passive safety systems, double containment, longer plant service life of 50 years with a capacity factor of 92%, longer irreplaceable components service life of 60 years, a 28.6% lower amount of concrete and metal, shorter construction time of 54 months, a Core Damage Frequency of 1x10-7/ year and lower liquid and solid wastes by 70% and 80% respectively. The presented paper includes a comparative analysis of technological and safety features, economic parameters and environmental impact of the AES-2006 design versus the other western advanced reactors. Since the Bushehr phase II NPP and several other NPPs are planning in Iran, such analysis would be of a great importance

  16. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  17. FIND: Standard Safety Analysis Report (GESSAR-238)

    International Nuclear Information System (INIS)

    This index is presented as a guide to microfiche items in Docket STN-50447, which was assigned to the BWR/6 STANDARD SAFETY ANALYSIS REPORT (GESSAR-238) submitted by General Electric Company, San Jose, California. The report describes and analyzes a standard BWR/6 boiling water reactor with a Mark III containment system designed for initial operation at approximately 3579 MW(t) with a net electrical output of approximately 1220 megawatts

  18. Conceptual design study of small sodium cooled reactors

    International Nuclear Information System (INIS)

    A conceptual design of various small metal fuel sodium cooled reactors has been studied in the feasibility study on commercialized fast breeder reactor cycle system. In FY2004 study, a 50 MWe power plant for remote places with a long life core without refueling and a 300 MWe modular reactor which pursues standardization for learning effect and reduction of capital risks. In the small reactor with a long life core, the reactor vessel is minimized without a permanent fuel handling system and the cooling system is simplified adopting 1 loop. The total mass of the reactor vessel and the cooling system is dramatically reduced and the concept has a potential to be an attractive power source for remote places. In the 300 MWe modular reactor, the cooling system adopts 1 loop and the ex-vessel fuel storage tank for spent fuels is eliminated adapting the in-vessel storage (IVS) which has a capacity for a 4 year storage. The reactor building is minimized without the ex-vessel storage. This concept has a potential to be an power source for key grids with modular constructions and a first plant with a small fuel cycle facility can demonstrate the metal fuel fast reactor cycle. (author)

  19. Development of mechanical design technology for integral reactor

    International Nuclear Information System (INIS)

    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were set up, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

  20. Development of mechanical design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Choi, Suhn; Kim, Kang Soo; Kim, Tae Wan; Jeong, Kyeong Hoon; Lee, Gyu Mahn

    1999-03-01

    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were setup, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

  1. Technical outline of a high temperature pool reactor with inherent passive safety features

    International Nuclear Information System (INIS)

    Many reactor designers world wide have successfully established technologies for very small reactors (less than 10 MWTH), and technologies for large power reactors (greater than 1000 MWTH), but have not developed small reactors (between 10 MWTH and 1000 MWth) which are safe, economic, and capable of meeting user technical, economic, and safety requirements. This is largely because the very small reactor technologies and the power reactor technologies are not amiable to safe and economic upsizing/downsizing. This paper postulates that new technologies, or novel combinations of existing technologies are necessary to the design of safe and economic small reactors. The paper then suggest a set of requirements that must be satisfied by a small reactor design, and defines a pool reactor that utilizes lead coolant and TRISO fuel which has the potential for meeting these requirements. This reactor, named LEADIR-PS, (an acronym for LEAD-cooled Integral Reactor, Passively Safe) incorporates the inherent safety features of the Modular High Temperature Gas Cooled Reactor (MWGR), while avoiding the cost of reactor and steam generator pressure vessels, and the safety concerns regarding pressure vessel rupture. This paper includes the description of a standard 200MW thermal reactor module based on this concept, called LEADIR-PS 200. (author)

  2. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  3. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  4. JNC viewpoint on fast reactor knowledge preservation

    International Nuclear Information System (INIS)

    JNC is undertaking a major program of research and development on liquid-metal cooled fast breeder reactors, which is fully supported by the government of Japan and the electrical utilities. Hence, the perspective of JNC on knowledge preservation is rather different from that of organizations where the fast reactor project has been scaled down or discontinued. Within JNC, there is a statutory obligation to preserve documentary records of the fast reactor project. Over time the method of archiving has changed from optical (microfilm, microfiche etc.) to digital storage. It is the long-term objective of JNC to convert all its records to digital format and make them available to staff over its intranet. JNC is also attempting to preserve 'human knowledge', that is, the expertise of staff who have been involved in the fast reactor project over a long period and who are now nearing retirement. Based on this information, two computerized systems are currently being constructed: one which records in,a readily accessible manner the background to key design decisions for the Monju plant; and a second which uses simple relationships between design parameters to aid designers understand the knock-on effects of design choices (joint project with Mitsubishi). To its partners in international cooperation - the US/DoE and the organizations of the Euro-Japan collaboration - JNC is proposing a joint approach to knowledge preservation and retrieval. The proposed concept, dubbed the International Super-Archive Network (ISAN), would make use of the standardized software the new technologies of the internet increase the mutual accessibility of fast reactor information. JNC considers it extremely important to reflect the lessons learnt from previous experience in the fast reactor field to the operation and maintenance of Monju and the design of future reactors. (author)

  5. WWER safety investigations on LR-0 reactor

    International Nuclear Information System (INIS)

    A set of the measurement needed for the WWER-440 and WWER-1000 reactor lifetime assessment, verification of the methods, codes and input cross section libraries for the WWER reactor pressure vessel exposure evaluation has been performed on the LR-0 experimental reactor. The WWER Mock-ups (engineering benchmarks) has been carried out on the reactor, with the aim to investigate differential neutron spectra for reactor dosimetry purposes. Critical experiments have also been performed to determine the perturbation of the fission density distribution caused by the WWER-440 control assembly. Such assembly, partially inserted in the core, has significant influence on the space power distribution. A wide research program for sub-criticality investigations of the spent nuclear fuel storage has been realized on the LR-0 reactor. A benchmark experiment is realized on the reactor in corresponding geometry for CASTOR 440/84 container for storage and transportation of spent fuel. Critical experiments with new fuel assemblies including various burnable absorbers and different enrichments are performed. A set of critical experiments is performed using the fuel assemblies with 3,6% and 4,4% enrichment, arranged in the WWER-440 type cores with various lattice pitch. The critical high of the moderator level and the moderator level coefficient of reactivity are measured and the effect of the fuel assembly, placed in a hexagonal tube of stainless steel containing boron absorber (ATABOR - STANDARD) is investigated. The obtained results are used for the validation of the codes (MCNP, KENO and SCALE) in the frame of the contract 'Burn-up credit implementation for the storage and transport containers of the spent fuel'. Combined neutron-gamma spectra measurements in the WWER-1000 Mock-up are carried out during 2001

  6. Final Report BW Sample Collection& Preparation Device

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, R P; Belgrader, P; Meyer, G; Benett, W J; Richards, J B; Hadley, D R; Stratton, P L; Milanovich, F P

    2002-01-31

    The objective of this project was to develop the technique needed to prepare a field collected sample for laboratory analysis and build a portable integrated biological detection instrument with new miniaturized and automated sample purification capabilities. The device will prepare bacterial spores, bacterial vegetative cells, and viral particles for PCR amplification.

  7. FBR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  8. Reactor water sampling device

    International Nuclear Information System (INIS)

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  9. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  10. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  11. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  12. Advancement of light water reactor technology

    International Nuclear Information System (INIS)

    The Japanese technology of light water reactors is based on the technology imported from abroad around 1970, and the experience has been accumulated by the construction, operation and repair of light water reactors as well as the countermeasures to various troubles, moreover, the improvement and standardization of light water reactors have been promoted. As the result, recently the high capacity ratio has been attained, and the LWR technology has firmly taken root in Japan. The Subcommittee for the Advancement of Light Water Reactor Technology of the Advisory Committee for Energy has examined the subjects of technical development and the way the development should be in order to decide the strategy to advance LWR technology, and drawn up the interim report. The change of situation around the LWRs in Japan and the necessity to advance the technology, the target of advancing LWR technology and the subjects of the technical development, the system for the technical development and the securement of fund, and international cooperation are reported. The subjects of development are the pursuit of higher reliability and economic efficiency, the extension of plant life, the improvement of repairability and the reduction of radiation exposure, the improvement of operational capability, the reduction of wastes, the techniques for reactor decommissioning and the diversified location. (Kako, I.)

  13. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  14. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  15. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

    Main research reactor safety parameters such as power density peaking factors, shutdown margin and temperature reactivity coefficients are treated. Reactor physics explanation of the parameters is given together with their application in safety evaluation performed as part of research reactor operation. Reactor calculations are presented as a method for their determination assuming use of widely available computer codes. (author)

  16. ASME initiatives for the support of new nuclear reactors

    International Nuclear Information System (INIS)

    With all the advances in nuclear power in the past years, the ASME Board on Nuclear Codes and Standards (BNCS) is investigating what it can do to prepare for anew generation of plants. The ASME BNCS has developed a Task Group to identify the needs of developers of new reactor designs. The goal of: this task group is to encourage the leadership of experts working on new reactor designs to communicate with code committees their requirements to ensure that all the appropriate needs relating to new reactor development are being addressed within the ASME codes and standards committee structures: New plant designs include the following types of reactors: advanced boiling water, advanced pressurized water, gas-cooled fast-spectrum, pebble bed modular, and high temperature gas-cooled. Many of these designs incorporate features that are not addressed by current nuclear codes and standards. Theses features include non-metallic materials, and configurations not amenable to current in-service inspection requirements. ASME would like to have the codes and standards in place for these designs during their development process, not only to apply a standard, but also to enhance acceptance of these new designs by users and regulators. The ASME BNCS is approaching the challenge by meeting with new reactor developers worldwide at their locations to discuss codes and standards needs. This paper will summarize initiatives already underway to address development of requirements for materials such as high temperature metallic materials, nuclear graphite, ceramics, carbon composites, and passive non-metallic components. Initiatives related to the use of probabilistic methods in design to support new reactors will also be discussed along with efforts to use risk-informed methods to develop inservice inspection requirements to support gas-cooled reactors. (authors)

  17. Improvement on Structure of Technical Standards by Ordinance of MOST

    International Nuclear Information System (INIS)

    Technical standards by legal term in Korea are the mandatory requirements which are prescribed in nuclear laws and Notices of the MOST (Minister of Science and Technology). Regulations on Technical Standards for Nuclear Reactor Facilities, etc. (Ministerial Ordinance No.31) prescribes the technical standards on the location, structure, installation, performance, operation, and quality assurance for the design, construction and operation of reactor facilities and nuclear fuel cycle facilities. Although the ordinance was wholly amended in July 2001, the technical standards by the ordinance should be improved in a way because they do not have separate and detail standards for different type of reactor, and still have some mutatis mutandis regulations and so on. Recently, globalization for the safety standards has been pursued. In this environment, technical standards have to be improved to meet the concept of the Atomic Energy Act and the international safety level. In this paper, the structure and components of the technical standards for reactor facilities were reviewed and evaluated comparing with foreign standards such as IAEA safety standards, and considering current regulatory position. Also, the components and contents of the new requirements were suggested

  18. Reactor de plasma

    OpenAIRE

    Erra Serrabasa, Pilar; Molina Mansilla, Ricardo; Beltrán Serra, Eric

    2008-01-01

    Reactor de plasma. Se trata de un reactor de plasma que puede trabajar en un amplio rango de presión, desde el vacío y presiones reducidas hasta la presión atmosférica y presiones superiores. Adicionalmente el reactor de plasma tiene la capacidad de regular otros parámetros importantes y permite su uso para el tratamiento de muestras de tipología muy diversa, como por ejemplo las de tamaño relativamente grande o de superficie rugosa.

  19. Integral nuclear reactor

    International Nuclear Information System (INIS)

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  20. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.