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Sample records for ce standard reactor

  1. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    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

  2. The national standards program for research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1977-01-01

    In 1970 a standards committee called ANS-15 was established by the American Nuclear Society (ANS) to prepare appropriate standards for research reactors. In addition, ANS acts as Secretariat for a national standards committee N17 which is responsible to the American National Standards Institute (ANSI) for the national consensus efforts for standards related to research reactors. To date ANS-15 has completed or is working on 14 standards covering all aspects of the operation of research reactors. Of the 11 research reactor standards submitted to the ANSI N17 Committee since its inception, six have been issued as National standards, and the remaining are still in the process of review. (author)

  3. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  4. Standards for safe operation of research reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The safety of research reactors is based on many factors such as suitable choice of location, design and construction according to the international standards, it also depends on well trained and qualified operational staff. These standards determine the responsibilities of all who are concerned with the research reactors safe operation, and who are responsible of all related activities in all the administrative and technical stages in a way that insures the safe operation of the reactor

  5. Materials surveillance program for C-E NSSS reactor vessels

    International Nuclear Information System (INIS)

    Koziol, J.J.

    1977-01-01

    Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

  6. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

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

  7. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

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

  8. Standards for reference reactor physics measurements

    International Nuclear Information System (INIS)

    Harris, D.R.; Cokinos, D.M.; Uotinen, V.

    1990-01-01

    Reactor physics analysis methods require experimental testing and confirmation over the range of practical reactor configurations and states. This range is somewhat limited by practical fuel types such as actinide oxides or carbides enclosed in metal cladding. On the other hand, this range continues to broaden because of the trend of using higher enrichment, if only slightly enriched, electric utility fuel. The need for experimental testing of the reactor physics analysis methods arises in part because of the continual broadening of the range of core designs, and in part because of the nature of the analysis methods. Reactor physics analyses are directed primarily at the determination of core reactivities and reaction rates, the former largely for reasons of reactor control, and the latter largely to ensure that material limitations are not violated. Errors in these analyses can be regarded as being from numerics, from the data base, and from human factors. For numerical, data base, and human factor reasons, then, it is prudent and customary to qualify reactor physical analysis methods against experiments. These experiments can be treated as being at low power or at high power, and each of these types is subject to an American National Standards Institute standard. The purpose of these standards is to aid in improving and maintaining adequate quality in reactor physics methods, and it is from this point of view that the standards are examined here

  9. Transformation of pristine and citrate-functionalized CeO2 nanoparticles in a laboratory-scale activated sludge reactor.

    Science.gov (United States)

    Barton, Lauren E; Auffan, Melanie; Bertrand, Marie; Barakat, Mohamed; Santaella, Catherine; Masion, Armand; Borschneck, Daniel; Olivi, Luca; Roche, Nicolas; Wiesner, Mark R; Bottero, Jean-Yves

    2014-07-01

    Engineered nanomaterials (ENMs) are used to enhance the properties of many manufactured products and technologies. Increased use of ENMs will inevitably lead to their release into the environment. An important route of exposure is through the waste stream, where ENMs will enter wastewater treatment plants (WWTPs), undergo transformations, and be discharged with treated effluent or biosolids. To better understand the fate of a common ENM in WWTPs, experiments with laboratory-scale activated sludge reactors and pristine and citrate-functionalized CeO2 nanoparticles (NPs) were conducted. Greater than 90% of the CeO2 introduced was observed to associate with biosolids. This association was accompanied by reduction of the Ce(IV) NPs to Ce(III). After 5 weeks in the reactor, 44 ± 4% reduction was observed for the pristine NPs and 31 ± 3% for the citrate-functionalized NPs, illustrating surface functionality dependence. Thermodynamic arguments suggest that the likely Ce(III) phase generated would be Ce2S3. This study indicates that the majority of CeO2 NPs (>90% by mass) entering WWTPs will be associated with the solid phase, and a significant portion will be present as Ce(III). At maximum, 10% of the CeO2 will remain in the effluent and be discharged as a Ce(IV) phase, governed by cerianite (CeO2).

  10. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  11. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    Moir, R.W.

    1978-01-01

    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

  12. 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.

  13. Advances in U.S. reactor physics standards

    International Nuclear Information System (INIS)

    Cokinos, Dimitrios

    2008-01-01

    The standards for Reactor Design, widely used in the nuclear industry, provide guidance and criteria for performing and validating a wide range of nuclear reactor calculations and measurements. Advances, over the past decades in reactor technology, nuclear data and infrastructure in the data handling field, led to major improvements in the development and application of reactor physics standards. A wide variety of reactor physics methods and techniques are being used by reactor physicists for the design and analysis of modern reactors. ANSI (American National Standards Institute) reactor physics standards, covering such areas as nuclear data, reactor design, startup testing, decay heat and fast neutron fluence in the pressure vessel, are summarized and discussed. These standards are regularly undergoing review to respond to an evolving nuclear technology and are being successfully used in the U.S and abroad contributing to improvements in reactor design, safe operation and quality assurance. An overview of the overall program of reactor physics standards is presented. New standards currently under development are also discussed. (authors)

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

    International Nuclear Information System (INIS)

    Richards, Wade J.

    1984-01-01

    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

  15. An overview of reactor physics standards: Past, present and future

    International Nuclear Information System (INIS)

    Cokinos, D.M.

    1992-07-01

    This report discusses for determining key static reactor physics parameters which have been developed by groups of experts (working groups) under the aegis of ANS-19, the ANS Reactor Physics Standards Committee. Following a series of sequential reviews, augmented by feedback from potential users, a proposed standard is brought into final form by the working group before it is adopted as a formal standard by the American National Standards Institute (ANSI); Reactor Physics standards are intended to provide guidance in the performance and qualification of complex sequences of reactor calculations and/or measurements and are regularly reviewed for possible updates and/or revisions. The reactor physics standards developed to date are listed and standards now being developed by the respective working groups are also provided

  16. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

  17. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The 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 steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  18. American National Standard: nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    1983-01-01

    This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include criteria for acceptance of evaluated nuclear data sets, criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  19. American National Standard nuclear data sets for reactor design calculations

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    A standard is presented which identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types

  20. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    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

  1. Overview of standards subcommittee 8, fissionable materials outside reactors

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1996-01-01

    The American Nuclear Society's Standards Subcommittee 8, titled open-quotes Fissionable Materials Outside Reactors,close quotes has worked for the past 35 yr to prepare and promote standards on nuclear criticality safety for the handling, processing, storing, and transportation of fissionable materials outside reactors. The reader is referred to the Transactions of the American Nuclear Society, Vols. 39 (1981) and 64 (1991), for previous papers associated with ANS-8 poster sessions. In addition to discussions on the then-current standards, the reader will find articles on working group efforts that never materialized into standards, such as proposed 8.13, open-quotes Use of the Solid-Angle Method in Nuclear Criticality Safety,close quotes and on applications and critiques of current standards. The paper by McLendon in Vol. 39 is particularly interesting as an overview of the early history of ANS-8 and its standards

  2. 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.

  3. Reference design for the standard mirror hybrid reactor

    International Nuclear Information System (INIS)

    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.

    1978-01-01

    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 ( 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

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

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2010-01-01

    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

  5. Standard irradiation facilities for use in TRIGA reactors

    International Nuclear Information System (INIS)

    Kolbasov, B.N.; Luse, R.A.

    1972-01-01

    The standard neutron irradiation facility (SNIP) was developed under IAEA and FAO co-ordinated research program for the standardization of neutron irradiation facilities for radiobiological research, resulting in the possibility to use fast neutrons from pool-type reactors for radiobiological studies. The studies include irradiation of seeds for crop improvement, of Drosophila for genetic studies, and of microorganisms for developing industrially useful mutants, as well as fundamental studies in radiation biology. The facilities, located in the six pool-type reactors (in Austria, Bulgaria, India, Philippines, Thailand and Taiwan), have been calibrated and utilized to compare the response to fast neutrons of barley seeds (variety Himalaya CI 000620) which were selected as a standard biological monitor by which to estimate neutron fluxes in different reactors. These comparative irradiation studies showed excellent agreement and reproducibility

  6. Outlines of revised regulation standards for experimental research reactors

    International Nuclear Information System (INIS)

    Hohara, Shinya

    2015-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Schulte, S.C.; Willke, T.L.; Young, J.R.

    1978-05-01

    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

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

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

    SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

  9. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    GRIFFIN, PATRICK J.

    1999-01-01

    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

  10. Piping benchmark problems for the ABB/CE System 80+ Standardized Plant

    International Nuclear Information System (INIS)

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1994-07-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the ABB/Combustion Engineering System 80+ Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the System 80+ standard design. It will be required that the combined license licensees demonstrate that their solution to these problems are in agreement with the benchmark problem set. The first System 80+ piping benchmark is a uniform support motion response spectrum solution for one section of the feedwater piping subjected to safe shutdown seismic loads. The second System 80+ piping benchmark is a time history solution for the feedwater piping subjected to the transient loading induced by a water hammer. The third System 80+ piping benchmark is a time history solution of the pressurizer surge line subjected to the accelerations induced by a main steam line pipe break. The System 80+ reactor is an advanced PWR type

  11. Non-standard interaction effects at reactor neutrino experiments

    International Nuclear Information System (INIS)

    Ohlsson, Tommy; Zhang, He

    2009-01-01

    We study non-standard interactions (NSIs) at reactor neutrino experiments, and in particular, the mimicking effects on θ 13 . We present generic formulas for oscillation probabilities including NSIs from sources and detectors. Instructive mappings between the fundamental leptonic mixing parameters and the effective leptonic mixing parameters are established. In addition, NSI corrections to the mixing angles θ 13 and θ 12 are discussed in detailed. Finally, we show that, even for a vanishing θ 13 , an oscillation phenomenon may still be observed in future short baseline reactor neutrino experiments, such as Double Chooz and Daya Bay, due to the existences of NSIs

  12. A proposed standard on medical isotope production in fission reactors

    International Nuclear Information System (INIS)

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-01-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  13. Influence of CeO2 NPs on biological phosphorus removal and bacterial community shifts in a sequencing batch biofilm reactor with the differential effects of molecular oxygen

    International Nuclear Information System (INIS)

    Xu, Yi; Wang, Chao; Hou, Jun; Wang, Peifang; You, Guoxiang; Miao, Lingzhan; Lv, Bowen; Yang, Yangyang

    2016-01-01

    The effects of CeO 2 nanoparticles (CeO 2 NPs) on a sequencing batch biofilm reactor (SBBR) with established biological phosphorus (P) removal were investigated from the processes of anaerobic P release and aerobic P uptake. At low concentration (0.1 mg/L), no significant impact was observed on total phosphorus (TP) removal after operating for 8 h. However, at a concentration of 20 mg/L, TP removal efficiency decreased from 83.68% to 55.88% and 16.76% when the CeO 2 NPs were added at the beginning of the anaerobic and aerobic periods, respectively. Further studies illustrated that the inhibition of the specific P release rate was caused by the reversible states of Ce 3+ and Ce 4+ , which inhibited the activity of exopolyphosphatase (PPX) and transformation of poly-β-hydoxyalkanoates (PHA) and glycogen, as well as the uptake of volatile fatty acids (VFAs). The decrease in the specific P uptake rate was mainly attributed to the significantly suppressed energy generation and decreased abundance of Burkholderia caused by excess reactive oxygen species. The removal of chemical oxygen demand (COD) was not influenced by CeO 2 NPs under aerobic conditions, due to the increased abundance of Acetobacter and Acidocella after exposure. The inhibitory effects of CeO 2 NPs with molecular oxygen were reduced after anaerobic exposure due to the enhanced particle size and the presence of Ce 3+ . - Highlights: • CeO 2 NPs (20 mg/L) had a notable toxicity effect on P removal in SBBR system. • The deteriorated SPRR was caused by the inhibited key enzyme activity (PPX). • The decreased SPUR was caused by the bacterial community shifts. • Ce ions converting and excess ROS generation are related toxicity mechanisms.

  14. International standardization of safety requirements for fast reactors

    International Nuclear Information System (INIS)

    2011-06-01

    Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

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

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1990-01-01

    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

  16. IAEA safety standards and approach to safety of advanced reactors

    International Nuclear Information System (INIS)

    Gasparini, M.

    2004-01-01

    The paper presents an overview of the IAEA safety standards including their overall structure and purpose. A detailed presentation is devoted to the general approach to safety that is embodied in the current safety requirements for the design of nuclear power plants. A safety approach is proposed for the future. This approach can be used as reference for a safe design, for safety assessment and for the preparation of the safety requirements. The method proposes an integration of deterministic and risk informed concepts in the general frame of a generalized concept of safety goals and defence in depth. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor including small and medium sized reactors with innovative safety features.(author)

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

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    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)

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

    International Nuclear Information System (INIS)

    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

  19. Non-Power Reactor Operator Licensing Examiner Standards

    International Nuclear Information System (INIS)

    1994-06-01

    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 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

  20. 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.

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

    International Nuclear Information System (INIS)

    Sliter, G.E.

    1997-01-01

    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

  2. Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    The reduction of air pollution has become an international concern over the last ten years because of increases in emissions from mobile and stationary sources. Among these sources, volatile organic compounds (VOC) represent a serious environmental problem, together with NOx, SOx and particulate...... VOC component of Diesel engine exhausts, over a wide range of temperatures. The entire reactor was thought as a highly porous catalytic filter for a possible application in a Diesel exhausts purification system. The porous reactor was used as a backbone for the infiltration of Co3O4 and Co3O4/CeO2...

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

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

    1983-01-01

    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)

  4. Comparison of standard fast reactor calculations (Baker model)

    Energy Technology Data Exchange (ETDEWEB)

    Voropaev, A I; Van' kov, A A; Tsybulya, A M

    1978-12-01

    Compared are standard fast reactor calculations performed at different laboratories using several nuclear data files: BNAB-70 and OSKAR-75 (the USSR), CARNAVAL-4 (France), FD-5 (Great Britain), KFK-INR (West Germany), ENDF/B4 (the USA). Three fuel compositions were chosen: (1) /sup 239/Pu and /sup 238/U; (2) /sup 239/Pu, /sup 238/U and fission products; (3) /sup 239/Pu, /sup 240/Pu, /sup 238/U and fission products. Medium temperature was 300K. The calculations have been conducted in the diffusion approximation. Data on critical masses and breeding ratios are tabulated. Discrepancies in the calculations of all the characteristics are small since all the countries possess practically the same nuclear data files.

  5. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Science.gov (United States)

    2013-12-09

    [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...: timothy.kolb@nrc.gov . SUPPLEMENTARY INFORMATION: I. Accessing Information and Submitting Comments A...

  6. 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...

  7. Fusion reactor design studies: standard unit costs and cost scaling rules

    International Nuclear Information System (INIS)

    Schulte, S.C.; Bickford, W.E.; Willingham, C.E.; Ghose, S.K.; Walker, M.G.

    1979-09-01

    This report establishes standard unit costs and scaling rules for estimating costs of material, equipment, land, and labor components used in magnetic confinement fusion reactor plant construction and operation. Use of the standard unit costs and scaling rules will add uniformity to cost estimates, and thus allow valid comparison of the economic characteristics of various reactor concepts

  8. Plasma-catalyst hybrid reactor with CeO2/γ-Al2O3 for benzene decomposition with synergetic effect and nano particle by-product reduction.

    Science.gov (United States)

    Mao, Lingai; Chen, Zhizong; Wu, Xinyue; Tang, Xiujuan; Yao, Shuiliang; Zhang, Xuming; Jiang, Boqiong; Han, Jingyi; Wu, Zuliang; Lu, Hao; Nozaki, Tomohiro

    2018-04-05

    A dielectric barrier discharge (DBD) catalyst hybrid reactor with CeO 2 /γ-Al 2 O 3 catalyst balls was investigated for benzene decomposition at atmospheric pressure and 30 °C. At an energy density of 37-40 J/L, benzene decomposition was as high as 92.5% when using the hybrid reactor with 5.0wt%CeO 2 /γ-Al 2 O 3 ; while it was 10%-20% when using a normal DBD reactor without a catalyst. Benzene decomposition using the hybrid reactor was almost the same as that using an O 3 catalyst reactor with the same CeO 2 /γ-Al 2 O 3 catalyst, indicating that O 3 plays a key role in the benzene decomposition. Fourier transform infrared spectroscopy analysis showed that O 3 adsorption on CeO 2 /γ-Al 2 O 3 promotes the production of adsorbed O 2 - and O 2 2‒ , which contribute benzene decomposition over heterogeneous catalysts. Nano particles as by-products (phenol and 1,4-benzoquinone) from benzene decomposition can be significantly reduced using the CeO 2 /γ-Al 2 O 3 catalyst. H 2 O inhibits benzene decomposition; however, it improves CO 2 selectivity. The deactivated CeO 2 /γ-Al 2 O 3 catalyst can be regenerated by performing discharges at 100 °C and 192-204 J/L. The decomposition mechanism of benzene over CeO 2 /γ-Al 2 O 3 catalyst was proposed. Copyright © 2017 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1981-11-01

    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

  10. 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 ...

  11. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Science.gov (United States)

    2013-09-30

    ... Design Certification and Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Design Certification and Operating Reactors.'' The NRC seeks comments on the proposed revised section of... subject): Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013...

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

    International Nuclear Information System (INIS)

    O'Dell, R.D.

    1977-09-01

    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

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

    International Nuclear Information System (INIS)

    2010-01-01

    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

  14. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

    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

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

    International Nuclear Information System (INIS)

    Ramana, M.V.; Saikawa, Eri

    2011-01-01

    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.

  16. Influence of CeO{sub 2} NPs on biological phosphorus removal and bacterial community shifts in a sequencing batch biofilm reactor with the differential effects of molecular oxygen

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yi; Wang, Chao [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China); Hou, Jun, E-mail: hhuhjyhj@126.com [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China); Wang, Peifang, E-mail: pfwang2005@hhu.edu.cn [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China); You, Guoxiang; Miao, Lingzhan; Lv, Bowen; Yang, Yangyang [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China)

    2016-11-15

    The effects of CeO{sub 2} nanoparticles (CeO{sub 2} NPs) on a sequencing batch biofilm reactor (SBBR) with established biological phosphorus (P) removal were investigated from the processes of anaerobic P release and aerobic P uptake. At low concentration (0.1 mg/L), no significant impact was observed on total phosphorus (TP) removal after operating for 8 h. However, at a concentration of 20 mg/L, TP removal efficiency decreased from 83.68% to 55.88% and 16.76% when the CeO{sub 2} NPs were added at the beginning of the anaerobic and aerobic periods, respectively. Further studies illustrated that the inhibition of the specific P release rate was caused by the reversible states of Ce{sup 3+} and Ce{sup 4+}, which inhibited the activity of exopolyphosphatase (PPX) and transformation of poly-β-hydoxyalkanoates (PHA) and glycogen, as well as the uptake of volatile fatty acids (VFAs). The decrease in the specific P uptake rate was mainly attributed to the significantly suppressed energy generation and decreased abundance of Burkholderia caused by excess reactive oxygen species. The removal of chemical oxygen demand (COD) was not influenced by CeO{sub 2} NPs under aerobic conditions, due to the increased abundance of Acetobacter and Acidocella after exposure. The inhibitory effects of CeO{sub 2} NPs with molecular oxygen were reduced after anaerobic exposure due to the enhanced particle size and the presence of Ce{sup 3+}. - Highlights: • CeO{sub 2} NPs (20 mg/L) had a notable toxicity effect on P removal in SBBR system. • The deteriorated SPRR was caused by the inhibited key enzyme activity (PPX). • The decreased SPUR was caused by the bacterial community shifts. • Ce ions converting and excess ROS generation are related toxicity mechanisms.

  17. Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

    Science.gov (United States)

    Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

    2018-08-01

    Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.

  18. Recent developments for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, N.; Nakamuria, K.; Morishita, M.; Shibamoto, H.; Nagashima, H.; Inoue, K.

    2005-01-01

    For realization of reliable and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute(JNC) and Japan Atomic Power Company(JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts through evaluation of their structural integrity, the research and development of 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)' is recognized as an essential theme. FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damages due to cyclic thermal loads. To evaluate these modes, three main developments are in progress. One is 'Refinement of Failure Criteria' for particular modes of FRs. Next is development of 'Guidelines for Inelastic Design Analysis' in order to predict elastic plastic and creep behaviors. Furthermore, efforts are being made toward preparing 'Guidelines for Thermal Load Modeling' for FR component design where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry of Japanese government. (authors)

  19. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

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

    International Nuclear Information System (INIS)

    Bottimore, R.R.

    1980-12-01

    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

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

    International Nuclear Information System (INIS)

    1997-01-01

    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

  2. Conversion and standardization of university reactor fuels using low-enrichment uranium - Options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    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. The U.S. Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the U.S. Department of Energy. (author)

  3. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    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. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab

  4. Synthesis of nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} by absorption of ammonia into water-in-oil microemulsion in a rotor–stator reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yingwen; Wang, Hongrun; Arowo, Moses; Sun, Baochang, E-mail: sunbc@mail.buct.edu.cn; Chen, Jianfeng; Shao, Lei, E-mail: shaol@mail.buct.edu.cn [Beijing University of Chemical Technology, State Key Laboratory of Organic–Inorganic Composites (China)

    2015-01-15

    A gas-microemulsion reaction precipitation method was employed to prepare nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} by absorption of NH{sub 3} into water-in-oil (W/O) microemulsion in a rotor–stator reactor . The effects of different operating conditions including final pH of the microemulsion, reaction temperature, initial Ce{sup 3+} and Zr{sup 4+} concentration, rotation speed, and gas–liquid volumetric ratio were investigated. Nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} with an average diameter of about 5.5 nm, a specific surface area of 215.6 m{sup 2}/g and a size distribution of 4–8 nm was obtained under the optimum operating conditions. The as-prepared nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} was loaded with Au to prepare nano-Au/Ce{sub 0.5}Zr{sub 0.5}O{sub 2} catalyst which was subsequently used for CO oxidation test. CO conversion rate reached 100 % at room temperature, indicating high catalytic activity of the nano-Au/Ce{sub 0.5}Zr{sub 0.5}O{sub 2} catalyst.

  5. Standard technical specifications for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    1979-08-01

    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

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

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1995-10-01

    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

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

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    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)

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

    International Nuclear Information System (INIS)

    Berthelemy, Michel; Escobar Rangel, Lina

    2013-01-01

    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)

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

    International Nuclear Information System (INIS)

    Virgilio, M.

    1979-07-01

    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

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

    International Nuclear Information System (INIS)

    1978-06-01

    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

  11. Standard technical specifications for combustion engineering pressurized water reactors

    International Nuclear Information System (INIS)

    1979-08-01

    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

  12. Standard technical specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Wagner, P.C.

    1979-07-01

    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

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

    International Nuclear Information System (INIS)

    Berthélemy, Michel; Escobar Rangel, Lina

    2015-01-01

    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

  14. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

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

    International Nuclear Information System (INIS)

    1983-10-01

    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

  16. Safety of reactors built according to earlier standards (WWER 440/V230 type)

    International Nuclear Information System (INIS)

    Misak, J.; Rohar, S.

    1995-01-01

    The problems of safety of WWER-440/V-230 type reactors are discussed, and the following conclusions are made. (1) The reactors have a very good operational record. (2) The reactors have serious design shortcomings, which should be eliminated by safety upgrading. Core damage frequency should be further reduced. (3) PSA methods constitute an appropriate tool for assessment of plant vulnerability to some initiating events and malfunctions, for prioritization of upgrading measures and for tolerability of deviations from current safety standards. (4) The most important safety merits, such as a large thermal inertia and low rupture probability, should be properly taken into account in the analysis. (5) Extensive safety upgrading is feasible and can lead to a considerable risk reduction. In certain circumstances such upgrading is the least expensive option even though the total cost is much higher than the initial plant construction cost. (6) Properly upgraded, the reactor units may be operable until better power resources are available within the country. (7) The existing gap between the technological and political judgements of nuclear safety should be reduced continuously by information exchange improvements. (8) A unified approach to nuclear safety should be adopted for all nuclear reactors (not just WWERs) built to earlier standards. 5 tabs., 1 fig

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

    International Nuclear Information System (INIS)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao; 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)

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

    International Nuclear Information System (INIS)

    1988-06-01

    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

  19. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

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

    Directory of Open Access Journals (Sweden)

    Yu-Feng Li

    2014-11-01

    Full Text Available 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.

  1. Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practice E 185. 1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties. 1.3 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. (See Note 1.)

  2. Progress report on the IAEA programme on the standardization of reactor dosimetry measurements

    International Nuclear Information System (INIS)

    Ertek, C.; Cross, B.; Chernyshev, V.

    1979-01-01

    This report briefly summarizes present activities, current status and procedures associated with neutron spectrum unfolding by activation technique within the IAEA programme on standardization of reactor radiation measurements. Experimental efforts and calculations related to unfolding are critically analyzed including the most recent techniques, interlaboratory cooperation, direct influence of recently measured cross-sections on the unfolded neutron flux density spectrum, re-evaluation of some cross-sections, neutron self-shielding factors and scattering effects. (author)

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

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    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 January 1981

  4. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  5. Nuclear reactors. Use of the protection system for non-safety purposes (International Electrotechnical Commission Standard Publication 639:1979)

    International Nuclear Information System (INIS)

    Stefanik, J.

    1996-01-01

    This standard applies to the protection system of a nuclear reactor and, more especially, to all interconnections between a reactor protection system (as defined and explained in International Electrotechnical Commission Publication 231 A, first supplement to Publication 231, General Principles of Nuclear Reactor Instrumentation) and all other systems and equipment not part of the protection system, except: a) the physical connection between sensors of the protection system and the physical variables that they monitor, such as for example, thermo wells, moderating medium for neutron sensors, etc.; b) the electrical connection between the protection system and the reactor control rods or other safety mechanism; c) the electrical and pneumatic connections to the power distribution system (mains) and pneumatic supplies that supply power to the protection system. Although many clauses relate to all reactor protection systems, this standard applies mainly to protection systems in nuclear power reactors

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

    International Nuclear Information System (INIS)

    Rico, N.

    1999-01-01

    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)

  7. Proposal of an ISO Standard: Classification of Transients and Accidents for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Chung, Bub Dong; Lee, Doo-Jeong; Kim, Jong In; Yoon, Ju Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, An Sup; Lee, Sang Yoon [Korea Electric Association, Seoul (Korea, Republic of)

    2016-05-15

    Classification of the events for a nuclear power plant is a fundamental basis for defining nuclear safety functions, safety systems performing those functions, and specific acceptance criteria for safety analyses. Presently, the approaches for the event classification adopted by the nuclear suppliers are different, which makes a nuclear technology trade barrier. The IAEA and WENRA are making efforts to establish general requirements or guidelines on the classification of either plant states or defence-in-depth levels for the design of nuclear power plants. However, the requirements and guidelines do not provide the details for practical application to various types of commercial PWRs. Recently, Korea proposed a new ISO standardisation project to develop a harmonized or consolidated international standard for classifying the events in PWRs and for defining (or imposing) the acceptance criteria for reactor design and/or radiation protection corresponding to each event class. This paper briefs the method with strategies for developing the standard, the current various practices of the PWR event classification and acceptance criteria developed or adopted by several organizations in USA and Europe, and a draft of the proposed standard. The proposed standard will affect all the relevant stakeholders such as reactor designers, vendors, suppliers, utilities, regulatory bodies, and publics of the leading countries in the area of nuclear industry as well as utilities, regulatory bodies, and publics of the newly entering (starting) countries. It is expected that all of the stakeholders will benefit from the proposed deliverable which provides an internationally harmonized standard for classifying the PWR events as follows: The reactor design bases for assuring safety and related technical information can be effectively communicated and shared among them resulting in enhancement of the global nuclear safety and fosterage of the global nuclear trade. The countries starting

  8. Standard practice for analysis and interpretation of physics dosimetry results for test reactors

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This practice describes the methodology summarized in Annex Al to be used in the analysis and interpretation of physics-dosimetry results from test reactors. This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods that are in various stages of completion (see Fig. 1). Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods. This practice is directed towards the development and application of physics-dosimetrymetallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practice E 853, Practice E 560, Matrix E 706(IE), Practice E 185, Matrix E 706(IG), Guide E 900, and Method E 646

  9. The conceptual design of the standard and the reduced fuel assemblies for an advanced research reactor

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Cho, Yeong Garp; Yoon, Doo Byung; Dan, Ho Jin; Chae, Hee Tack; Park, Cheol

    2005-01-01

    HANARO (Hi-flux Advanced Neutron Application Reactor), is an open-tank-in-pool type research reactor with a thermal power of 30MW. The HANARO has been operating at Korea Atomic Energy Research Institute since 1995. Based on the technical experiences in design and operation for the HANARO, the design of an Advanced Research Reactor (ARR) was launched by KAERI in 2002. The final goal of the project is to develop a new and advanced research reactor model which is superior in safety and economical aspects. This paper summarizes the design improvements of the conceptually designed standard fuel assembly based on the analysis results for the nuclear physics. It includes also the design of the reduced fuel assembly in conjunction with the flow tube as the fuel channel and the guide of the absorber rod. In the near future, the feasibility of the conceptually designed fuel assemblies of the ARR will be verified by investigating the dynamic and the thermal behaviors of the fuel assembly submerged in coolant

  10. Research and development issues for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

    2003-01-01

    For realization of safe and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts and validate their structural integrity, the research and development of 'Fast Reactor Structural Design Standard (FDS)' is recognized as an essential theme. FDS considers general characteristics of FRs and design needs for their rationalization. Three main subjects were settled in research and development issues of FDS. One is rationalization of failure criteria' taking characteristic design conditions into account. Next is development of 'a guideline on inelastic analysis for design' in order to predict elastic plastic and creep behaviours of high temperature components. Furthermore, efforts are being made toward preparing a guideline on thermal loads modeling' for FR component design where thermal loads are dominant. (author)

  11. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    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)

  13. Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

    A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

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

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-10-01

    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

  15. A General Small-Scale Reactor To Enable Standardization and Acceleration of Photocatalytic Reactions.

    Science.gov (United States)

    Le, Chi Chip; Wismer, Michael K; Shi, Zhi-Cai; Zhang, Rui; Conway, Donald V; Li, Guoqing; Vachal, Petr; Davies, Ian W; MacMillan, David W C

    2017-06-28

    Photocatalysis for organic synthesis has experienced an exponential growth in the past 10 years. However, the variety of experimental procedures that have been reported to perform photon-based catalyst excitation has hampered the establishment of general protocols to convert visible light into chemical energy. To address this issue, we have designed an integrated photoreactor for enhanced photon capture and catalyst excitation. Moreover, the evaluation of this new reactor in eight photocatalytic transformations that are widely employed in medicinal chemistry settings has confirmed significant performance advantages of this optimized design while enabling a standardized protocol.

  16. On the actual controlling of standards concerning the 'fast breeder reactor'

    International Nuclear Information System (INIS)

    Rengeling, H.W.

    1978-01-01

    If the decision of the OVG Muenster to present the case to the Federal Constitutional Court asking whether Article 7 of the atomic energy law corresponds to the constitution or not, as far as the article allows the licensing of a fast breeder reactor, two problems arise: The legal question in how far the actual controlling of standards is to be preceeded by a statement of facts given by the court of first instance, and the problem behind concerning the responsibility of decision. The Federal Constitutional Court should accept the responsibility of decision to be borne by it. (orig.) [de

  17. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  18. DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made.

  19. DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium

    International Nuclear Information System (INIS)

    1994-01-01

    Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made

  20. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  1. 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

    ... Standard Review Plan, Section 1.0 on Introduction and Interfaces AGENCY: Nuclear Regulatory Commission (NRC... Revision 2 to Standard Review Plan (SRP), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110110573). The Office of New Reactors (NRO...

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

    International Nuclear Information System (INIS)

    Bates, E.F.; Grimes, B.K.; Ramos, S.L.

    1982-05-01

    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

  3. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

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

    International Nuclear Information System (INIS)

    Fichtner, N.; Becker, K.; Bashir, M.

    1977-01-01

    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

  5. Standard deviation of local tallies in global Monte Carlo calculation of nuclear reactor core

    International Nuclear Information System (INIS)

    Ueki, Taro

    2010-01-01

    Time series methodology has been studied to assess the feasibility of statistical error estimation in the continuous space and energy Monte Carlo calculation of the three-dimensional whole reactor core. The noise propagation was examined and the fluctuation of track length tallies for local fission rate and power has been formally shown to be represented by the autoregressive moving average process of orders p and p-1 [ARMA(p,p-1)], where p is an integer larger than or equal to two. Therefore, ARMA(p,p-1) fitting was applied to the real standard deviation estimation of the power of fuel assemblies at particular heights. Numerical results indicate that straightforward ARMA(3,2) fitting is promising, but a stability issue must be resolved toward the incorporation in the distributed version of production Monte Carlo codes. The same numerical results reveal that the average performance of ARMA(3,2) fitting is equivalent to that of the batch method with a batch size larger than 100 and smaller than 200 cycles for a 1,100 MWe pressurized water reactor. (author)

  6. A computer program for accident calculations of a standard pressurized water reactor

    International Nuclear Information System (INIS)

    Keutner, H.

    1979-01-01

    In this computer program the dynamic of a standard pressurized water reactor should be realized by both circulation loops with all important components. All important phenomena are taken into consideration, which appear for calculation of disturbances in order to state a realistic process for some minutes after a disturbance or a desired change of condition. In order to optimize the computer time simplifications are introduced in the statement of a differential-algebraic equalization system such that all important effects are taken into consideration. The model analysis starts from the heat production of the fuel rod via cladding material to the cooling medium water and considers the delay time from the core to the steam generator. Alternations of the cooling medium pressure as well as the different temperatures in the primary loop influence the pressuring system - the pressurizer - which is realized by a water and a steam zone with saturated and superheated steam respectively saturated and undercooled water with injection, heating and blow-down devices. The bilance of the steam generator to the secondary loop realizes the process engineering devices. Thereby the control regulation of the steam pressure and the reactor performance is realized. (orig.) [de

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

    International Nuclear Information System (INIS)

    Marston, T.U.; Layman, W.H.; Bockhold, G. Jr.

    1992-01-01

    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

  8. Catalytic Oxidation of CO and Soot over Ce-Zr-Pr Mixed Oxides Synthesized in a Multi-Inlet Vortex Reactor: Effect of Structural Defects on the Catalytic Activity.

    Science.gov (United States)

    Bensaid, Samir; Piumetti, Marco; Novara, Chiara; Giorgis, Fabrizio; Chiodoni, Angelica; Russo, Nunzio; Fino, Debora

    2016-12-01

    In the present work, ceria, ceria-zirconia (Ce = 80 at.%, Zr = 20 at.%), ceria praseodymia (Ce = 80 at.%, Pr = 20 at.%) and ceria-zirconia-praseodymia catalysts (Ce = 80 at.%, Zr = 10 at.% and Pr = 10 at.%) have been prepared by the multi-inlet vortex reactor (MIVR). For each set of samples, two inlet flow rates have been used during the synthesis (namely, 2 ml min -1 , and 20 ml min -1 ) in order to obtain different particle sizes. Catalytic activity of the prepared materials has been investigated for CO and soot oxidation reactions. As a result, when the catalysts exhibit similar crystallite sizes (in the 7.7-8.8 nm range), it is possible to observe a direct correlation between the O v /F 2g vibrational band intensity ratios and the catalytic performance for the CO oxidation. This means that structural (superficial) defects play a key role for this process. The incorporation of Zr and Pr species into the ceria lattice increases the population of structural defects, as measured by Raman spectroscopy, according to the order: CeO 2  oxidation activity for these catalysts, in contrast with nanostructured ones (e.g., Ce-Zr-O nanopolyhedra, Ce-Pr-O nanocubes) described elsewhere (Andana et al. Appl. Catal. B 197: 125-137, 2016; Piumetti et al., Appl Catal B 180: 271-282, 2016).

  9. Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs; and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-70). 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E 706) (1, 5, 13, 48, 49). In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units. Note 1—(Figure 1 is deleted in the latest update. The user is refered to Master Matrix E 706 for the latest figure of the standards interconnectivity). 1.3 This practice is restri...

  10. Reference and standard benchmark field consensus fission yields for U.S. reactor dosimetry programs

    International Nuclear Information System (INIS)

    Gilliam, D.M.; Helmer, R.G.; Greenwood, R.C.; Rogers, J.W.; Heinrich, R.R.; Popek, R.J.; Kellogg, L.S.; Lippincott, E.P.; Hansen, G.E.; Zimmer, W.H.

    1977-01-01

    Measured fission product yields are reported for three benchmark neutron fields--the BIG-10 fast critical assembly at Los Alamos, the CFRMF fast neutron cavity at INEL, and the thermal column of the NBS Research Reactor. These measurements were carried out by participants in the Interlaboratory LMFBR Reaction Rates (ILRR) program. Fission product generation rates were determined by post-irradiation analysis of gamma-ray emission from fission activation foils. The gamma counting was performed by Ge(Li) spectrometry at INEL, ANL, and HEDL; the sample sent to INEL was also analyzed by NaI(Tl) spectrometry for Ba-140 content. The fission rates were determined by means of the NBS Double Fission Ionization Chamber using thin deposits of each of the fissionable isotopes. Four fissionable isotopes were included in the fast neutron field measurements; these were U-235, U-238, Pu-239, and Np-237. Only U-235 was included in the thermal neutron yield measurements. For the fast neutron fields, consensus yields were determined for three fission product isotopes--Zr-95, Ru-103, and Ba-140. For these fission product isotopes, a separately activated foil was analyzed by each of the three gamma counting laboratories. The experimental standard deviation of the three independent results was typically +- 1.5%. For the thermal neutron field, a consensus value for the Cs-137 yield was also obtained. Subsidiary fission yields are also reported for other isotopes which were studied less intensively (usually by only one of the participating laboratories). Comparisons with EBR-II fast reactor yields from destructive analysis and with ENDF/B recommended values are given

  11. 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.

  12. 75 FR 69709 - Office of New Reactors; Notice of Availability of the Final Staff Guidance; Standard Review Plan...

    Science.gov (United States)

    2010-11-15

    ... the Final Staff Guidance; Standard Review Plan, Section 13.6.6, Revision 0 on Cyber Security Plan... Reports for Nuclear Power Plants,'' Section 13.6.6, Revision 0 on ``Cyber Security Plan'' (Agencywide... reviews to amendments to licenses for operating reactors or for activities associated with review of...

  13. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  14. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  15. INAA of human and animal whole blood samples by short term reactor irradiation. [Au, Br, Cd, Ce, Cl, Cr, Cu, Fe, Hg, K, La, Mn, Na, P, Sc, Zn

    Energy Technology Data Exchange (ETDEWEB)

    Samudralwar, D L; Garg, A N

    1986-11-03

    Instrumental neutron activation analysis was employed for the determination of 15 major, minor and trace elements in human and animal blood samples. Dry whole blood samples along with NBS and IAEA standards were irradiated for 5 min, 1 h, 5 h and 10 h with reactor thermal neutrons and counted using high resolution ..gamma..-spectrometry at successive intervals. Data for a new IAEA proposed CRM Mixed Human Diet (H-9) is reported. 28 references, 4 tables.

  16. Hermetic cable penetrations for containments of nuclear power reactors meet high safety standards

    International Nuclear Information System (INIS)

    Kusserow, J.; Gurr, W.; Pflug, H.

    1985-05-01

    Different types of cable penetrations for containments of nuclear power reactors have been developed and fabricated in the GDR. The technical parameters achieved are in accordance with the radiation protection requirements

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

    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{sup -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

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

    International Nuclear Information System (INIS)

    1990-04-01

    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

  19. Different magnetic behaviour of the Kondo compounds Al3Ce and Al11Ce3

    International Nuclear Information System (INIS)

    Benoit, A.; Flouquet, J.; Palleau, J.; Buevoz, J.L.

    1979-08-01

    Neutron diffraction experiments on the Al 3 Ce and Al 11 Ce 3 compounds have been performed on the multidetector of the I.L.L. high flux reactor. No magnetic structure has been detected on the Al 3 Ce compound down to 20 mK. This confirms the non magnetic ground state of Al 3 Ce. For Al 11 Ce 3 , two magnetic structures have been observed: a ferromagnetic one at 4.2 K and an antiferromagnetic one at 2 K. The antiferromagnetic structure, which corresponds to a propagation vector (0,0,1/3), implies a strong reduction of the magnetic moment of determined sites; this reflects the Kondo character of the compounds

  20. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  1. The effect of economic growth, oil prices, and the benefits of reactor standardization: Duration of nuclear power plant construction revisited

    International Nuclear Information System (INIS)

    Csereklyei, Zsuzsanna; Thurner, Paul W.; Bauer, Alexander; Küchenhoff, Helmut

    2016-01-01

    The profitability of nuclear power plant investment is largely determined by the construction duration, which directly impacts discounted cash flows, debt and interest payments, as well as variable costs, such as labor. This paper analyzes the key drivers of construction duration using survival models. We focus especially on the strategic expectation formation of private and public utilities engaging in such highly risky megaprojects. Using a balanced dataset of explanatory variables and the IAEA/PRIS dataset of reactor construction starts between 1950 and 2013 we find that the expectation of rising oil prices and higher economic growth, along with the higher per capita GDP of a country tend to reduce the time needed to grid connection. We also identify the reactor models with the fastest construction duration. - Highlights: • We find that higher future economic growth speeds up nuclear reactor construction. • Higher national capacity (measured by income per capita) results in faster projects. • Higher oil prices during construction lead to faster construction times. • Reactor standardization may result in faster building times.

  2. 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...

  3. 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...

  4. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  5. Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

    1996-01-01

    A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

  6. Uncertainty determination of analysis of Ti, V, Cl, Ce, Cr, Cs, Sc, Co, Fe and Ca in solid samples by INAA method using standard addition according to ISO - guide 17025

    International Nuclear Information System (INIS)

    Sumining; Agus Taftazani

    2003-01-01

    Uncertainty of analysis of Ti, V, Cl, Ce, Cr, Cs, Sc, Co, Fe and Ca in solid samples by INAA (/instrumental Neutron Activation Analysis) method using comparative technique and standard addition have been carried out at INAA laboratory of P3TM BATAN. The calculation of Ti have been presented as the example. Uncertainty sources of INAA are sampling, sample and standard preparation, irradiation and counting. Sample were come from IAEA (International Atomic Energy Agency) which had ready for analyzed therefore only for sample and standard preparation, irradiation and counting factors were determined. Analysis were done by relative technique, that sample and standard were irradiated together in same capsule therefore irradiation time, neutron flux, irradiation geometry and isotopic properties. will be eliminated. Uncertainty of counting factors were covering radioactivity decay during the counting, pulse losses caused by random counting, counting geometry, and counting rate. Relative technique makes the uncertainty come from counting time for sample and standard that was settled by same counting equipment can be neglected. Uncertainty of counting geometry and thickness of uranium was not detected so there is no contribution come from The fission product. Variation of fuel target nuclides number didn't occurred because the combustion was not occurred during irradiation, and analytical results were not influenced by the chemical status. (author)

  7. Multielemental analysis of IAEA intercomparison standard Hay Powder, V-10 and some edible plant leaves by neutron activation. [Br,Ce,Cl,Cr,Cu,Fe,Ga,Hg,K,La,Mn,Mo,Na,P,Sc,Zn

    Energy Technology Data Exchange (ETDEWEB)

    Samudralwar, D L; Wankhade, H K; Garg, A N

    1987-12-01

    Instrumental neutron activation analysis was employed for the multielement determination of an IAEA intercomparison standard Hay Powder, V-10 and some edible plant leaves consumed in India. The samples were irradiated with thermal neutrons at a flux approx. = 10/sup 12/ n x cm/sup -2/ x s/sup -1/ in a reactor for 5 minutes, 1, 2, 5, 10 and 15 hours and counted by high resolution ..gamma..-ray spectrometry. Nearly 18 elements were determined. Good agreement is observed for most of the elements in several NBS standards and the proposed CRM V-10. Some edible vegetable plant leaves were also analyzed. (author) 32 refs.; 3 tabs.

  8. 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)

    1983-09-01

    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

  9. Axial stability of VVER-1000 reactor with control with minimum standard deviation

    International Nuclear Information System (INIS)

    Afanas'ev, A.M.; Torlin, B.Z.

    1980-01-01

    Results are given of investigations on the stability of a reactor which has, in addition to an automatic controller, a height distribution regulator (HDR) based on an auxiliary control rod (CR) or a special shortened absorption rod (SAR). The HDR was controlled by using either a special ionization chamber (IC), generating an imbalance signal which sets the CR in motion, or two ionization chambers whose difference signal causes a displacement of the SAR. Since data from numerous pickups can be used to control the height field of the VVER-1000, it is of interest to analyze how this would affect the stability of the reactor. The analysis was carried out with the improved IRINA programs. 11 refs

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

    International Nuclear Information System (INIS)

    Gauché, François

    2013-01-01

    • 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

  11. Standardized dose factors for dose calculations - 1982 SRP reactor safety analysis report tritium, iodine, and noble gases

    International Nuclear Information System (INIS)

    Pillinger, W.L.; Marter, W.L.

    1982-01-01

    Standardized dose constants are recommended for calculation of offsite doses in the 1982 SRP Reactor Safety Analysis Report (SAR). Dose constants are proposed for inhalation of tritium and radioiodines and for submersion in a semi-infinite cloud of radioiodines and noble gases. The proposed constants, based on ICRP2 methodology for internal dose and methodology recommended by the US Nuclear Regulatory Commission for external dose, are compatible with dose calculational methods used at the Savannah River Plant and Savannah River Laboratory for normal releases of radioactivity. 8 references

  12. Analyses and results from standard surveillance programmes of WWER 440/V-213C reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Falcnik, M; Brumovsky, M; Pav, T [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    In Czech and Slovak republics, six units of WWER 440/C type reactors are monitored by surveillance specimens programmes; the specimens are determined for static tensile testing, impact notch toughness testing and fracture toughness evaluation. Results of mechanical properties of these specimens after irradiation in intervals between 1 and 5 years of operation, are summarized and discussed with respect to the effect of individual heats and welded joints, radiation embrittlement, and annealing recovery. (authors). 3 refs., 11 figs., 2 tabs.

  13. Geometry modeling for SAM-CE Monte Carlo calculations

    International Nuclear Information System (INIS)

    Steinberg, H.A.; Troubetzkoy, E.S.

    1980-01-01

    Three geometry packages have been developed and incorporated into SAM-CE, for representing in three dimensions the transport medium. These are combinatorial geometry - a general (non-lattice) system, complex combinatorial geometry - a very general system with lattice capability, and special reactor geometry - a special purpose system for light water reactor geometries. Their different attributes are described

  14. Survey of legal aspects, regulations, standards and guidelines applicable to radioactive waste management of the Brazilian Multipurpose Reactor - RMB

    International Nuclear Information System (INIS)

    Salvetti, T.C.; Marumo, J.T.

    2017-01-01

    In Brazil, the Brazilian Nuclear Energy Commission (CNEN) and Brazilian Institute of Environment and Renewable Natural Resources (IBAMA) are the agencies responsible for the execution, regulation and control of nuclear and environmental policies, respectively. Such regulatory activities are very comprehensive (IBAMA) or too specific (CNEN), revealing other aspects that would, also, need to be observed so that the management could be carried out efficiently (quality) and effectively (safety), including the three governmental administrative levels: Federal, State and Municipal. In addition to laws, regulations, decrees and resolutions, there are also national and international standards and guides that provide guidelines for structuring the current management and the use of best regulatory practices. The Brazilian Multipurpose Reactor Enterprise (RMB) is a CNEN project, complying with a Multi-Year Plan of the Brazilian Ministry of Planning, Development and Management (MPDG). The Enterprise is being developed under the responsibility of the Directorate of Research and Development - DPD of CNEN and will have a facility for treatment and initial temporary storage of the radioactive waste generated by the operation of the research reactor and the activities carried out in the associated laboratories. The RMB will be built in the city of IPERÓ, located in the state of São Paulo, near ARAMAR Experimental Center of the Brazilian Navy. This work aims to present the research results regarding the various aspects that regulate, legislate and standardize the practices proposed to the Radioactive Waste Management of the RMB project. (author)

  15. Survey of legal aspects, regulations, standards and guidelines applicable to radioactive waste management of the Brazilian Multipurpose Reactor - RMB

    Energy Technology Data Exchange (ETDEWEB)

    Salvetti, T.C.; Marumo, J.T., E-mail: salvetti@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    In Brazil, the Brazilian Nuclear Energy Commission (CNEN) and Brazilian Institute of Environment and Renewable Natural Resources (IBAMA) are the agencies responsible for the execution, regulation and control of nuclear and environmental policies, respectively. Such regulatory activities are very comprehensive (IBAMA) or too specific (CNEN), revealing other aspects that would, also, need to be observed so that the management could be carried out efficiently (quality) and effectively (safety), including the three governmental administrative levels: Federal, State and Municipal. In addition to laws, regulations, decrees and resolutions, there are also national and international standards and guides that provide guidelines for structuring the current management and the use of best regulatory practices. The Brazilian Multipurpose Reactor Enterprise (RMB) is a CNEN project, complying with a Multi-Year Plan of the Brazilian Ministry of Planning, Development and Management (MPDG). The Enterprise is being developed under the responsibility of the Directorate of Research and Development - DPD of CNEN and will have a facility for treatment and initial temporary storage of the radioactive waste generated by the operation of the research reactor and the activities carried out in the associated laboratories. The RMB will be built in the city of IPERÓ, located in the state of São Paulo, near ARAMAR Experimental Center of the Brazilian Navy. This work aims to present the research results regarding the various aspects that regulate, legislate and standardize the practices proposed to the Radioactive Waste Management of the RMB project. (author)

  16. 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)

    1982-08-01

    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

  17. Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

    Science.gov (United States)

    Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

    2017-10-01

    Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

  18. 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)

    Bernard, H.

    1984-03-01

    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

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

    International Nuclear Information System (INIS)

    Meinke, W.W.; Essig, T.H.

    1991-04-01

    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

  20. Development of a standard for calculation and measurement of the moderator temperature coefficient of reactivity in water-moderated power reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.; Hall, R.A.; Lancaster, D.B.; Young, E.H.; Gavin, P.H.; Robertson, S.T.

    1998-01-01

    The contents of ANS 19.11, the standard for ''Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity in Water-Moderated Power Reactors,'' are described. The standard addresses the calculation of the moderator temperature coefficient (MTC) both at standby conditions and at power. In addition, it describes several methods for the measurement of the at-power MTC and assesses their relative advantages and disadvantages. Finally, it specifies a minimum set of documentation requirements for compliance with the standard

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

    International Nuclear Information System (INIS)

    Montoya R, E.

    1995-01-01

    It has been characterized and standardized an irradiation of the RP-10 Research Nuclear Reactor for use of the K 0 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 K 0 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

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

    International Nuclear Information System (INIS)

    Meinke, W.W.; Essig, T.H.

    1991-04-01

    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

  3. The application of the k0-standardization method at the TRIGA Mark II reactor, Ljubljana, Slovenia

    International Nuclear Information System (INIS)

    Jacimovic, Radojko; Benedik, Ljudmila; Stegnar, Peter; Smodis, Borut

    2002-01-01

    The k 0 -standardization method of neutron activation analysis (k 0 -NAA) was launched in the 1970s and since then continuously developed. Nowadays, k 0 -NAA became widespread as a practical analytical tool used to analyse different sample matrices. At the Jozef Stefan Institute (IJS), the KAYZERO/SOLCOI software package has been introduced for data processing after extensive testing and comparison with other available programs. In the process of validation of the software a suite of natural matrix reference materials (RMs) were used. Five certified reference materials (CRMs) from the Institute for Reference Materials and Measurements (IRMM), two standard reference materials (SRMs) from the National Institute of Standards and Technology (NIST), three RMs from the International Atomic Energy Agency (IAEA) and one RM from IJS were analysed. Altogether, results for ten elements in inorganic matrices and twenty-one elements in organic matrices, obtained by k 0 -instrumental neutron activation analysis (k 0 -INAA), were compared to certified values. The results obtained show good agreement with certified or assigned values except for Fe and U in inorganic matrices, and Al and Cr in organic matrices. (author)

  4. Measure of the efficiency of a long counter of Hanson's type and use of this counter for the survey of the slow neutrons coming from the reactor of Chatillon; Mesure de l'efficacite d'un long compteur du type Hanson et utilisation de ce compteur a l'etude des neutrons lents sortant de la pile de Chatillon

    Energy Technology Data Exchange (ETDEWEB)

    Barloutaud, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1953-07-01

    A detection device of fast neutrons of efficiency almost independent of the energy of the neutrons has been achieved. It efficiency has been measured in absolute value for groups of neutrons of different energies. This device allowed to get some indications on the energy composition of the neutrons leaving from the reactor of Chatillon. (author) [French] Un dispositif de detection de neutrons rapides d'efficacite pratiquement independante de l'energie des neutrons a ete realise. Son efficacite a ete mesuree en valeur absolue pour des groupes de neutrons de diverses energies. Ce dispositif a permis obtenir quelques indications sur la composition energetique des neutrons sortant de la pile de Chatillon. (auteur)

  5. A base to standardize data processing of cadmium ratio RCd and thermal neutron flux measurements on reactor

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1993-08-01

    The cadmium ratio R Cd and thermal neutron flux are usually measured in a reactor. But its data process is rather complex. The results from same measured data differ by different existing process methods. The purpose of this work is to standardize data processing in R Cd and thermal neutron flux measurements. A natural choice for this purpose is to derive a R Cd formula based on standard average thermal activation cross section and resonance integral and to define related parameters or factors that provide an unique base for comparison between different measurements in laboratories. The parameters or factors include E c , F m , F m ' and G th ' in thermal energy region due to upper truncated Maxwellian distribution and E Cd , F Cd , G r and S r in intermediate energy region. They are the function of multiple variables. The Au foil is used as an example to demonstrate their behaviors by chosen figures and tables which provide for practical data process by hand. The work also discusses limitation of R Cd measurement in terms of so called available and optimum region and notes that Co and Mn foils have a much wider available region among Au, In, Mn, W and Co, the commonly used detector foils

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

    International Nuclear Information System (INIS)

    Orvini, E.; Speziali, M.; Salvini, A.; Herborg, C.

    2002-01-01

    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

  7. 75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on...

    Science.gov (United States)

    2010-06-24

    ... Standard Review Plan Section 13.6.1, Revision 1 on Physical Security--Combined License and Operating...), Section 13.6.1 on ``Physical Security--Combined License and Operating Reactors,'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100350158). The Office of Nuclear Security and...

  8. Generating material strength standards of aluminum alloys for research reactors. Pt. 1. Yield strength values Sy and tensile strength values Su

    International Nuclear Information System (INIS)

    Tsuji, H.; Miya, K.

    1995-01-01

    Aluminum alloys are frequently used as structural materials for research reactors. The material strength standards, however, such as the yield strength values (S y ), the tensile strength values (S u ) and the design fatigue curve -which are needed to use aluminum alloys as structural materials in ''design by analysis'' - for those materials have not been determined yet. Hence, a series of material tests was performed and the results were statistically analyzed with the aim of generating these material strength standards. This paper, the first in a series on material strength standards of aluminum alloys, describes the aspects of the tensile properties of the standards. The draft standards were compared with MITI no. 501 as well as with the ASME codes, and the trend of the available data also was examined. It was revealed that the draft proposal could be adopted as the material strength standards, and that the values of the draft standards at and above 150 C for A6061-T6 and A6063-T6 could be applied only to the reactor operating conditions III and IV. Also the draft standards have already been adopted in the Science and Technology Agency regulatory guide (standards for structural design of nuclear research plants). (orig.)

  9. Investigating the capability of ToF-SIMS to determine the oxidation state of Ce

    Science.gov (United States)

    Seed Ahmed, H. A. A.; Swart, H. C.; Kroon, R. E.

    2018-04-01

    The capability of time of flight secondary ion mass spectrometry (ToF-SIMS) to determine the oxidation state of Ce ions doped in a phosphor was investigated. Two samples of SiO2:Ce (4 mol%) with known Ce3+/Ce4+ relative concentrations were subjected to ToF-SIMS measurements. The spectra were very similar and no significant differences in the relative peak intensities were observed that would readily allow one to distinguish Ce3+ from Ce4+. Although ToF-SIMS was therefore not useful to distinguish the charge state of Ce ions doped in this phosphor material, the idea in principle was also tested on two other samples, namely CeF3 and CeF4 These contain Ce as part of the host (i.e. much higher concentration) and are fluorides, which is significant because ToF-SIMS has previously been reported to be able to distinguish Eu2+ from Eu3+ in Eu doped Sr5(PO4)3F phosphor. The spectrum of CeF4 contained a small peak related to Ce4+ which was not observed in the CeF3 spectrum, yet the peak related to the Ce3+ ions was found to be much more intense in the spectrum of CeF4 than CeF3, showing that the ToF-SIMS signals cannot be directly interpreted as retaining the charge state of the ions in the original material. Nevertheless, the significant differences in the Ce-related peaks in the ToF-SIMS spectra from CeF3 and CeF4 show that the charge state of Ce may be distinguished. This study shows that while in principle ToF-SIMS may be used to distinguish the charge state of Ce ions, this depends on the sample and it would not be easy to interpret the spectra without a standard or reference.

  10. Applications of a lead pile coupled with fast reactor core of Yayoi as an intermediate energy neutron standard field

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakazawa, Masaharu; Sekiguchi, Akira; Wakabayashi, Hiroaki.

    1976-10-01

    Intermediate neutron column of YAYOI reactor is here evaluated as an intermediate energy neutron standard field which provides a base of the measurements of various reaction rates in that energy region, including detector calibration and Doppler coefficient determination. The experiments were performed using YAYOI's core as a fast neutron source by coupling with the large lead pile, which is a 160 ton's octagon of 2.5 m high and with a thickness of about 2.5 m face to face distance. Spatial variation of the neutron flux in the lead pile was estimated by gold activation foils, and the neutron spectrum by sandwich foils, a helium-3 proportional counter and a proton recoil counter. The calculated results were obtained using one and two- dimensional discrete ordinate code, ANISN and TWOTRAN II. Through comparison of experiment with calculation, it became clear that the neutron field at the central block has simple energy spectrum and stable spatial distribution of the neutron flux, the absolute of which was 5.0 x 10 4 (n/cm 2 /sec/Watt) at the representative energy of 1 KeV. The energy spectrum of the position and the spatial dependent neutron flux in the lead pile are both represented by the semiempirical formula, which must be useful both for evaluation of experimental data and for future applications. (auth.)

  11. Development of an accident consequence assessment code for evaluating site suitability of light- and heavy-water reactors based on the Korean Technical standards

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Won Tae; Jeong, Hae Sung; Jeong, Hyo Joon; Kil, A Reum; Kim, Eun Han; Han, Moon Hee [Nuclear Environment Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

  12. Recovery Ce from Ce - TBP Used Oxalic Acid

    International Nuclear Information System (INIS)

    Purwani, MV; Subagiono, R.; Suyanti

    2007-01-01

    Recovery or stripping Ce from Ce - TBP product of monazite sand used oxalic acid. Ce - TBP as organic phase and oxalic acid as aqueous phase and as strong precipitant compound to precipitate metal element. The stripping product as Ce - oxalic precipitate. The influence parameter were percentage of oxalic acid, volume ratio of Ce-TBP with oxalic acid, time and rate of stripping. At stripping of 25 ml Ce - TBP used oxalic acid, the optimum condition were achieve at using 5% oxalic acid, volume ratio of Ce - TBP : 5% oxalic acid = 1 : 1, time of stripping 7.5 minute and rate of stripping 150 rpm. At the optimum condition was obtained the recovery efficiency was 100%. (author)

  13. Molten Salt Reactor Experiment Facility (Building 7503) standards/requirements identification document adherence assessment plan at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-02-01

    This is the Phase 2 (adherence) assessment plan for the Building 7503 Molten Salt Reactor Experiment (MSRE) Facility standards/requirements identification document (S/RID). This document outlines the activities to be conducted from FY 1996 through FY 1998 to ensure that the standards and requirements identified in the MSRE S/RID are being implemented properly. This plan is required in accordance with the Department of Energy Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 90-2, November 9, 1994, Attachment 1A. This plan addresses the major aspects of the adherence assessment and will be consistent with Energy Systems procedure QA-2. 7 ''Surveillances.''

  14. CE APPROVAL IN ELECTRICAL HOUSEHOLD APPLIANCES AND A CASE STUDY

    Directory of Open Access Journals (Sweden)

    Nazmi EKREN

    2009-01-01

    Full Text Available Due to the reason for rapidly developing technology, increasing competition medium, and awareness of the consumers, nowadays, the exigency of production with good quality has gained more and more significance. Certification of the quality and safety of the products to the consumers is compulsory in terms of producers. There are some documents to certify safety of the products. One of them is CE certificate. In this paper, basic information about CE mark is given and CE standards and tests required for electrical household appliances are mentioned. As an application, one of an electrical household appliance, toaster grill is treated and examined. To obtain CE certificate for toaster grill, required tests are made according to EN60335-2-9 and CE certificate is obtained.

  15. 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.

  16. Superconductivity and anomalous normal state in the CePd2Si2/CeNi2Ge2 system

    International Nuclear Information System (INIS)

    Grosche, F.M.; Lister, S.J.S.; Carter, F.V.; Saxena, S.S.; Haselwimmer, R.K.W.; Mathur, N.D.; Julian, S.R.; Lonzarich, G.G.

    1997-01-01

    The unconventional nonmagnetic metal CeNi 2 Ge 2 is characterised at ambient pressure by temperature dependences of the specific heat and of the resistivity which deviate strongly from standard Fermi-liquid predictions and are reminiscent of the behaviour observed in its sibling system CePd 2 Si 2 above the critical pressure at which magnetic order is suppressed. We have explored the CePd 2 Si 2 /CeNi 2 Ge 2 phase diagram in a series of resistivity measurements under high hydrostatic pressure, p. At p>15 kbar, a new superconducting transition appears below 220 mK in CeNi 2 Ge 2 and shifts to higher temperatures with increasing pressure, reaching ∝400 mK at p∝26 kbar. (orig.)

  17. Electro-regeneration of Ce(IV) in real spent Cr-etching solutions

    International Nuclear Information System (INIS)

    Chen, Te-San; Huang, Kuo-Lin

    2013-01-01

    Highlights: • An electrochemical process is used to regenerate Ce(IV) in real (hazardous) spent TFT-LCD Cr-etching solutions. • The Ce(IV) yield on tested anodes was in order BDD > Pt > DSA. • A Neosepta CMX separator was better than Nafion ones to be used in the process. • The activation energy on Pt was 10.7 kJ/mol. • The obtained parameters are useful to design reactors for 100% Ce(IV) regeneration in real spent Cr-etching solutions. -- Abstract: This paper presents the electro-regeneration of Ce(IV) in real (hazardous) spent thin-film transistor liquid-crystal display (TFT-LCD) Cr-etching solutions. In addition to Ce(III) > Ce(IV) in diffusivity, a quasi-reversible behavior of Ce(III)/Ce(IV) was observed at both boron-doped diamond (BDD) and Pt disk electrodes. The Ce(IV) yield on Pt increased with increasing current density, and the best current efficiency (CE) was obtained at 2 A/2.25 cm 2 . The performance in terms of Ce(IV) yield and CE of tested anodes was in order BDD > Pt > dimensional stable anode (DSA). At 2 A/2.25 cm 2 on Pt and 40 °C for 90 min, the Ce(IV) yield, CE and apparent rate constant (k) for Ce(III) oxidation were 81.4%, 21.8% and 3.17 × 10 −4 s −1 , respectively. With the increase of temperature, the Ce(IV) yield, CE, and k increased (activation energy = 10.7 kJ/mol), but the specific electricity consumption decreased. The Neosepta CMX membrane was more suitable than Nafion-117 and Nafion-212 to be used as the separator of the Ce(IV) regeneration process. The obtained parameters are useful to design divided batch reactors for the Ce(IV) electro-regeneration in real spent Cr-etching solutions

  18. CE-BEMS

    DEFF Research Database (Denmark)

    Mohamed, Nader; Lazarova-Molnar, Sanja; Al-Jaroodi, Jameela

    2016-01-01

    and costs savings in smart buildings significantly depend on the monitoring and control methods used in the installed BEMS. This paper proposes a Cloud-Enabled BEMS (CE-BEMS) for Smart Buildings. This system can utilize cloud computing to provide enhanced management mechanisms and features for energy...... savings in smart buildings. This system is connected to the cloud to have access to a number of advanced cloud-based services to enhance energy management in smart buildings. In this paper, we discuss the current limitations of BEMS, the conceptual design of the proposed system, and the advantages......Energy consumption in smart buildings is monitored and controlled using Building Energy Management Systems (BEMS). A BEMS provides a set of methods to monitor and control a building's energy needs while maintaining a good quality of living in all of the building's spaces. Energy efficiency...

  19. CE and nanomaterials - Part II: Nanomaterials in CE.

    Science.gov (United States)

    Adam, Vojtech; Vaculovicova, Marketa

    2017-10-01

    The scope of this two-part review is to summarize publications dealing with CE and nanomaterials together. This topic can be viewed from two broad perspectives, and this article is trying to highlight these two approaches: (i) CE of nanomaterials, and (ii) nanomaterials in CE. The second part aims at summarization of publications dealing with application of nanomaterials for enhancement of CE performance either in terms of increasing the separation resolution or for improvement of the detection. To increase the resolution, nanomaterials are employed as either surface modification of the capillary wall forming open tubular column or as additives to the separation electrolyte resulting in a pseudostationary phase. Moreover, nanomaterials have proven to be very beneficial for increasing also the sensitivity of detection employed in CE or even they enable the detection (e.g., fluorescent tags of nonfluorescent molecules). © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  20. Supplementary control points for reactor shutdown without access to the main control room (International Electrotechnical Commission Standard Publication 965:1989)

    International Nuclear Information System (INIS)

    Kubalek, J.; Hajek, B.

    1993-01-01

    This standard establishes the requirements for supplementary Control Points provided to enable the operating staff to shut down the reactor and maintain the plant in a safe shut-down condition when the main control room is no longer available. This standard covers the functional selection, design and organization of the man/machine interface. It also establishes requirements for procedures which systematically verify and validate the functional design of supplementary control points. The requirements reflect the application of human engineering principles as they apply to man/machine interface. This standard does not cover special emergency response centres (e.g. a Technical Support Centre). It also does not include the detailed equipment design. Unavailability of the main control room controls due to intentionally man-induced events is not considered

  1. CE Challenges : Work to Do

    NARCIS (Netherlands)

    Stjepandic, J; Verhagen, W.J.C.; Wognum, P.M.

    2015-01-01

    CE has been used for more than two decades now. Despite many successes and advantages, there are still many challenges to be addressed. These challenges are both technical and organisational. In the paper we will address the current challenges of CE. Many challenges

  2. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  3. 75 FR 6413 - Office of New Reactors; Proposed Revision to Standard Review Plan, Section 14.3.12 on Physical...

    Science.gov (United States)

    2010-02-09

    ... Standard Review Plan, Section 14.3.12 on Physical Security Hardware Inspections, Tests, Analyses, and.... SUMMARY: The NRC is soliciting public comment on NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 1 to Standard Review Plan (SRP...

  4. Variegated operation of MAPS reactors after enmasse' coolant channel replacement: a tale-tell signature of high standard fuel bundle production quality

    International Nuclear Information System (INIS)

    Jena, J.K.; Sahu, J.K.; Arularasan, V.; Sivagurnathan, D.; Rathakrishnan, S.; Ramamurthy, K.

    2009-01-01

    After the Enmasse' Coolant Channel Replacement (EMCCR) of both the reactors of Madras Atomic Power Station (MAPS), they have put up a good performance, as far as core integrity is considered. This is a tale-tell signature of the high quality of the fuel bundles manufactured by Nuclear Fuel Complex (NFC), Hyderabad. Both the reactor cores have been loaded with various types of fuel bundles viz. Natural Uranium (NU), Depleted Uranium (DU), and Deeply Depleted Uranium (DDU) and were operated at different power level with different flux configuration at different stages of operation. Even around 1026 low burn up bundle (<2500 MWD/TeU) were transferred from MAPS-1 to MAPS-2, first time in the history of PHWRS. During all such variegated operations, the Primary Heat Transport (PHT) system 131 I activity, which is synonymous with the core integrity, was maintaining low for most of the reactor operation period. However, recently a low burn up fuel bundle failure has been observed in MAPS-1. Even though the overall failure rate is very low, the cause of such failure needs to be ascertained for taking appropriate action to maintain the high standards of quality in the manufacturing process of the fuel bundles. (author)

  5. TRX and UO2 criticality benchmarks with SAM-CE

    International Nuclear Information System (INIS)

    Beer, M.; Troubetzkoy, E.S.; Lichtenstein, H.; Rose, P.F.

    1980-01-01

    A set of thermal reactor benchmark calculations with SAM-CE which have been conducted at both MAGI and at BNL are described. Their purpose was both validation of the SAM-CE reactor eigenvalue capability developed by MAGI and a substantial contribution to the data testing of both ENDF/B-IV and ENDF/B-V libraries. This experience also resulted in increased calculational efficiency of the code and an example is given. The benchmark analysis included the TRX-1 infinite cell using both ENDF/B-IV and ENDF/B-V cross section sets and calculations using ENDF/B-IV of the TRX-1 full core and TRX-2 cell. BAPL-UO2-1 calculations were conducted for the cell using both ENDF/B-IV and ENDF/B-V and for the full core with ENDF/B-V

  6. 75 FR 29588 - Office of New Reactors: Proposed NUREG-0800; Standard Review Plan Section 13.6.6, Draft Revision...

    Science.gov (United States)

    2010-05-26

    ...; Standard Review Plan Section 13.6.6, Draft Revision 0 on Cyber Security Plan AGENCY: Nuclear Regulatory... Plants,'' on a proposed Standard Review Plan (SRP) Section 13.6.6 on ``Cyber Security Plan'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML093560837). The Office of Nuclear Security...

  7. 76 FR 31381 - Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

    Science.gov (United States)

    2011-05-31

    ... Standard Review Plan; Section 8.1 on Electric Power--Introduction AGENCY: U.S. Nuclear Regulatory...,'' on a proposed Revision 4 to Standard Review Plan (SRP), Section 8.1 on ``Electric Power--Introduction,'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML111180542). The previous version of...

  8. 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...

  9. Influence of Ce Doping on the Electrical and Optical Properties of TiO2 and Its Photocatalytic Activity for the Degradation of Remazol Brilliant Blue R

    Directory of Open Access Journals (Sweden)

    Aisha Malik

    2013-01-01

    Full Text Available Nanocrystalline TiO2 particles doped with different concentrations of Cerium (Ce, 1–10% have been synthesized using sol-gel method. The prepared particles were characterized by standard analytical techniques such as X-ray diffraction (XRD, FTIR and Scanning Electron Microscopy (SEM, and Transmission Electron Microscopy (TEM. The XRD analysis shows no change in crystal structure of TiO2 after doping with different concentrations of Ce, which indicates the single-phase polycrystalline material. The SEM analysis shows the partial crystalline nature of undoped, and doped TiO2 and TEM analysis shows the particle sizes were in the range of 9–14 nm in size. The a.c. analysis shows that the dielectric constant ε and dielectric loss tan δ decrease with the increase in frequency. The dielectric property decreases with the increase in dopant concentration. It is also observed that the impedance increases with an increase in dopant concentration. The photocatalytic activity of the synthesized particles (Ce-doped TiO2 with dopant concentration of 9% (Ce showed the highest photocatalytic activity for the degradation of the dye derivative Remazol Brilliant Blue R in an immersion well photochemical reactor with 500 W halogen linear lamp in the presence of atmospheric oxygen.

  10. 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...

  11. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  12. Oscillatory bromate-oxalic acid-Ce-acetone-sulfuric acid reaction, in CSTR

    International Nuclear Information System (INIS)

    Pereira, Janaina A.M.; Faria, Roberto B.

    2004-01-01

    Periodic oscillations were observed for the first time, in a CSTR, in the system bromate-oxalic acid-Ce(IV)-acetone-sulfuric acid, in a CSTR. A reaction between Ce(IV) and acetone, until now not described in the literature and occurring before the addition of the reagents to the reactor, was identified as a decisive factor for the appearing of the regular oscillations. (author)

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

    International Nuclear Information System (INIS)

    Hervouet, C.; Ranval, W.; Parozzi, F.; Eusebi, M.

    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 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

  14. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    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

  15. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1978-01-01

    We have carried out conceptual design studies 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 fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. 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

  16. Privatized multipurpose reactor initiative

    International Nuclear Information System (INIS)

    Davis, G.A.

    1995-01-01

    ABB Combustion Engineering (ABB CE) and seven other companies have submitted a plan to the DOE for deploying a multipurpose reactor at the Savannah River Plant. The facility would consume excess plutonium as fuel, irradiate tritium producing targets, and generate electricity. The plan proposes to establish a consortium that would privately finance and own two System 80+ nuclear units and a mixed oxide fuel fabrication facility

  17. Lattice dynamics of γ--Ce

    International Nuclear Information System (INIS)

    Gould, T.A.

    1978-08-01

    The phonon and magnetic measurements described in the thesis produced the following significant results concerning the lattice dynamical and magnetic properties of γ-Ce. The phonon spectrum is relatively soft, which is consistent with results obtained for CeSn 3 . The L [110] and T [111] branches of the dispersion curve are anomalous. The C 11 and C 44 elastic constants are quite close in value. No discrete magnetic excitations were observed. The magnetic scattering is qualitatively similar to the results from Ce 0 . 74 Th 0 . 26 , however, GAMMA/sub Ce/ less than GAMMA/sub Ce-Th/. The various lattice dynamical and magnetic similarities among γ-Ce, CeSn 3 , and Ce 0 . 74 Th 0 . 26 are mixed valence compounds. Therefore, a complete theoretical description of the observed properties of Ce and its compounds may provide a basis for understanding a whole class of mixed valence materials

  18. KOBİ Finansal Raporlama Standardının Türkiye’deki Muhasebe Uygulamalarına Getire-ceği Değişiklikler

    OpenAIRE

    CENGİZ, Hülya

    2014-01-01

    Contribution of small and medium sized entities to employment are significant in country`s economy due to their high number of existence and the added value they create. However their importance in the economy they experience difficulties in supplying a loan, expanding to international markets and in growing. One of these difficulties is about financial reporting. International Accounting Standards Board published the International Financial Reporting Standard for Small and Medium sized Entit...

  19. Heat capacity measurement of CeNbO4(s)

    International Nuclear Information System (INIS)

    Bhojane, S.M.; Kulkarni, Jayanthi; Kulkarni, S.G.

    2012-01-01

    Molar heat capacity of CeNbO 4 (s) was determined using differential scanning calorimeter in the temperature range of 550 to 900 K. The molar heat capacity values were least squares analysed and the dependence of molar heat capacity with temperature for CeNbO 4 (s) can be given as, J K -1 mol -1 = 94.7320 + 0.0852T-1.6073 x 10 6 T -2 (550≤T(K)≤900) Cerium is commonly used as an inactive analogue to plutonium; also it is an important fission product with moderate yield. Various Nb alloys are used as cladding material in nuclear industry. Hosts of thermodynamic data are needed to understand the various phenomena that occur in a nuclear reactor. In the present study, the molar heat capacity of CeNbO 4 (s) has been determined using high temperature differential scanning calorimeter in temperature range 550 to 900 K. This is one of the important compounds in the ternary system of Ce-Nb-O

  20. GAGG:ce single crystalline films: New perspective scintillators for electron detection in SEM

    International Nuclear Information System (INIS)

    Bok, Jan; Lalinský, Ondřej; Hanuš, Martin; Onderišinová, Zuzana; Kelar, Jakub; Kučera, Miroslav

    2016-01-01

    Single crystal scintillators are frequently used for electron detection in scanning electron microscopy (SEM). We report gadolinium aluminum gallium garnet (GAGG:Ce) single crystalline films as a new perspective scintillators for the SEM. For the first time, the epitaxial garnet films were used in a practical application: the GAGG:Ce scintillator was incorporated into a SEM scintillation electron detector and it showed improved image quality. In order to prove the GAGG:Ce quality accurately, the scintillation properties were examined using electron beam excitation and compared with frequently used scintillators in the SEM. The results demonstrate excellent emission efficiency of the GAGG:Ce single crystalline films together with their very fast scintillation decay useful for demanding SEM applications. - Highlights: • First practical application of epitaxial garnet films demonstrated in SEM. • Improved image quality of SEM equipped with GAGG:Ce single crystalline thin film scintillator. • Scintillation properties of GAGG:Ce films compared with standard bulk crystal scintillators.

  1. Determination of buckling in the IPEN/MB-01 Reactor in cylindrical configuration

    Energy Technology Data Exchange (ETDEWEB)

    Purgato, Rafael Turrini; Bitelli, Ulysses d' Utra; Aredes, Vitor Ottoni; Silva, Alexandre F. Povoa da; Santos, Diogo Feliciano dos; Mura, Luis Felipe L.; Jerez, Rogerio, E-mail: rtpurgato@ipen.br, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    One of the key parameters in reactor physics is the buckling of a reactor core. It is related to important parameters such as reaction rates, nuclear power operation, fuel burning, among others. In a critical reactor, the buckling depends on the geometric and material characteristics of the reactor core. This paper presents the results of experimental buckling in the reactor IPEN/MB-01 nuclear reactor in its cylindrical configuration with 28 fuel rods along its diameter. The IPEN/MB-01 is a zero power reactor designed to operate at a maximum power of 100 watts, it is a versatile nuclear facility which allows the simulation of all the characteristics of a large nuclear power reactor and ideal for this type of measurement. We conducted a mapping of neutron flux inside the reactor and thereby determined the total buckling of the cylindrical configuration. The reactor was operated for an hour. Then, the activation of the fuel rods was measured by gamma spectrometry on a rod scanner HPGe detector. We analyzed the gamma photons of the {sup 239}Np (276,6 keV) for neutron capture and the {sup 143}Ce (293,3 keV) for fission on both {sup 238}U and {sup 235}U, respectively. We analyzed the axial and radial directions. Other measurements were performed using wires and gold foils in the radial and axial directions of the reactor core. The results showed that the cylindrical configuration compared to standard rectangular configuration of the IPEN/MB-01 reactor has a higher neutron economy, since in this configuration there is less leakage of neutrons. The Buckling Total obtained from the three methods was 95.84 ± 2.67 m{sup -2}. (author)

  2. Magnetic behaviour of cerium in Ce2 Sn5 and Ce3 Sn7, surstructures of Ce Sn3

    International Nuclear Information System (INIS)

    Stunault, A.

    1988-07-01

    The compound studied, Ce 2 Sn 5 and Ce 3 Sn 7 are both orthorhombic, surstructure of cubic Ce Sn 3 . Magnetic susceptibility measurements show in both compounds an antiferromagnetic order at low temperature and magnetization shows a high anisotropy. Magnetization densities are determined by polarized neutron diffraction. The cerium site which has two Ce atoms as nearest neighbourgs carries all the magnetism in both structures. For Ce 2 Sn 5 moments are directed as the high magnetization axis and structure is modulated. Ce 3 Sn 7 presents a simple antiferromagnetic order but moment are directed as low magnetization axis. Various transitions towards a ferromagnetic order are presented. Results are interpreted by measuring the difference between energy levels of crystalline field. A model of crystalline field and isotrope exchange agrees well with Ce 3 Sn 7 , but for Ce 2 Sn 7 it is necessary to reduce the magnetic moment which is typical of the Kondo effect [fr

  3. Reactor dosimetry integral reaction rate data in LMFBR Benchmark and standard neutron fields: status, accuracy and implications

    International Nuclear Information System (INIS)

    Fabry, A.; Ceulemans, H.; Vandeplas, P.; McElroy, W.N.; Lippincott, E.P.

    1977-01-01

    This paper provides conclusions that may be drawn regarding the consistency and accuracy of dosimetry cross-section files on the basis of integral reaction rate data measured in U.S. and European benchmark and standard neutron fields. In a discussion of the major experimental facilities CFRMF (Idaho Falls), BIGTEN (Los Alamos), ΣΣ (Mol, Bucharest), NISUS (London), TAPIRO (Roma), FISSION SPECTRA (NBS, Mol, PTB), attention is paid to quantifying the sensitivity of computed integral data relative to the presently evaluated accuracy of the various neutron spectral distributions. The status of available integral data is reviewed and the assigned uncertainties are appraised, including experience gained by interlaboratory comparisons. For all reactions studied and for the various neutron fields, the measured integral data are compared to the ones computed from the ENDF/B-IV and the SAND-II dosimetry cross-section libraries as well as to some other differential data in relevant cases. This comparison, together with the proposed sensitivity and accuracy assessments, is used, whenever possible, to establish how well the best cross-sections evaluated on the basis of differential measurements (category I dosimetry reactions) are reliable in terms of integral reaction rates prediction and, for those reactions for which discrepancies are indicated, in which energy range it is presumed that additional differential measurements might help. For the other reactions (category II), the inconsistencies and trends are examined. The need for further integral measurements and interlaboratory comparisons is also considered

  4. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

    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 [sr

  5. Homoleptic Ce(III) and Ce(IV) Nitroxide Complexes: Significant Stabilization of the 4+ Oxidation State

    Energy Technology Data Exchange (ETDEWEB)

    Bogart, Justin A.; Lewis, Andrew J.; Medling, Scott A.; Piro, Nicholas A.; Carroll, Patrick J.; Booth, Corwin H.; Schelter, Eric J.

    2014-06-25

    Electrochemical experiments performed on the complex Ce-IV[2-((BuNO)-Bu-t)py](4), where [2-((BuNO)-Bu-t)py](-) = N-tert-butyl-N-2-pyridylnitroxide, indicate a 2.51 V stabilization of the 4+ oxidation state of Ce compared to [(Bu4N)-Bu-n](2)[Ce(NO3)(6)] in acetonitrile and a 2.95 V stabilization compared to the standard potential for the ion under aqueous conditions. Density functional theory calculations suggest that this preference for the higher oxidation state is a result of the tetrakis(nitroxide) ligand framework at the Ce cation, which allows for effective electron donation into, and partial covalent overlap with, vacant 4f orbitals with delta symmetry. The results speak to the behavior of CeO2 and related solid solutions in oxygen uptake and transport applications, in particular an inherent local character of bonding that stabilizes the 4+ oxidation state. The results indicate a cerium(IV) complex that has been stabilized to an unprecedented degree through tuning of its ligand-field environment.

  6. Superdeformed bands in 130Ce

    International Nuclear Information System (INIS)

    Paul, E.S.; Semple, A.T.; Boston, A.J.; Joss, D.T.; Nolan, P.J.; Shepherd, S.L.

    1997-01-01

    Four superdeformed bands have been assigned to 130 Ce following a high-statistics γ-ray study using the EUROGAM II spectrometer. The strongest band exhibits two distinct backbends which, in one scenario, may be interpreted as crossings between high-j N = 6 neutron orbitals (νi 13/2 ) and low-j N = 4 orbitals (νd 3/2 ) in an unpaired system. (author)

  7. Near term feasibility of nuclear reactor for sea-water desalting: coupling of standard condensing nuclear power stations to low grade heat multieffect distillation plants

    International Nuclear Information System (INIS)

    Adar, J.; Manor, S.; Schaal, M.

    1977-01-01

    Commercial nuclear power reactors exist only in standard sizes and designs. No large nuclear back-pressure turbines are available today. Therefore, near term large scale nuclear desalination plants must be tailored to the NSSS sizes and available turbines and not the contrary. Standard condensing nuclear turbines could operate continuously with a back-presure of up to 5-7'' Hg (depending on the supplier). It means that they can exhaust huge amounts of steam at 56 0 C - 64 0 C with a loss of electricity production of 6% - 10% when compared to 2 1/2'' Hg normal condensing pressure. The horizontal aluminium tube multi-effect distillation process developed by ''Israel Desalination Engineering'' Ltd. is very suitable for the use of such low-grade heat: 4 to 9 effects can operate within these temperature ranges. A special flash-chamber constitutes a positive barrier against any possible contamination being carried over by the steam exhausted from the turbine to the desalination plant. Flow sheets, heat and mass balances have been prepared for two standard sizes of NSSS and turbines (1882sup(Mwth) and 2785sup(Mwth)), two ''back-pressures'' (5 1/2'' and 7'' Hg), and corresponding desalination plants. Only standard equipment is being used in the steam and electricity producing plant. The desalination plant consists of 6 to 12 parallel double lines, each of them similar to a large prototype now being designed and which is going to be coupled to an old fossil power station. Water production varies between 50 and 123 sup(us MGD) and water cost between 23 and 36 sup(cents)/M 3 . Total energy requirements of the desalination plant represent only 19 to 50% of the total water cost as against 75% for a single purpose plant. Costs are based on actual bids for the power plant and actual estimates for the desalination prototype. The operation is designed to be flexible so that the power plant can be operated either in conjunction with the desalination plant, or as a single purpose

  8. Examination of fast reactor fuels, FBR analytical quality assurance standards and methods, and analytical methods development: irradiation tests. Progress report, April 1--June 30, 1976, and FY 1976

    International Nuclear Information System (INIS)

    Baker, R.D.

    1976-08-01

    Characterization of unirradiated and irradiated LMFBR fuels by analytical chemistry methods will continue, and additional methods will be modified and mechanized for hot cell application. Macro- and microexaminations will be made on fuel and cladding using the shielded electron microprobe, emission spectrograph, radiochemistry, gamma scanner, mass spectrometers, and other analytical facilities. New capabilities will be developed in gamma scanning, analyses to assess spatial distributions of fuel and fission products, mass spectrometric measurements of burnup and fission gas constituents and other chemical analyses. Microstructural analyses of unirradiated and irradiated materials will continue using optical and electron microscopy and autoradiographic and x-ray techniques. Analytical quality assurance standards tasks are designed to assure the quality of the chemical characterizations necessary to evaluate reactor components relative to specifications. Tasks include: (1) the preparation and distribution of calibration materials and quality control samples for use in quality assurance surveillance programs, (2) the development of and the guidance in the use of quality assurance programs for sampling and analysis, (3) the development of improved methods of analysis, and (4) the preparation of continuously updated analytical method manuals. Reliable analytical methods development for the measurement of burnup, oxygen-to-metal (O/M) ratio, and various gases in irradiated fuels is described

  9. R-ES-ONAN--CE

    Indian Academy of Sciences (India)

    Mathematics, Engineering: Turbulence, stability of fluid flows, two phase flow, thermal science, biochemical engineering, polymer engineering, chemical reactor analysis, transport phenomena, interfacial phenomena, process metallurgy, physi- cal metallurgy, ceramics, composite materials and computer science or ...

  10. Magnetic behaviour of new Ce compounds

    Energy Technology Data Exchange (ETDEWEB)

    Sampathkumaran, E V [Tata Inst. of Fundamental Research, Bombay (India); Mallik, R [Tata Inst. of Fundamental Research, Bombay (India)

    1996-07-01

    We report initial results of our investigation on the magnetic behaviour of some new Ce compounds. The compounds, CeIr{sub 2}B{sub 2}C and CeIr{sub 2}Ge{sub 2}, do not appear to exhibit bulk magnetic ordering down to 2 K. The alloys, Ce{sub 2}Pd{sub 2}In and Ce{sub 2}Cu{sub 2}In, order magnetically below 4 and 6 K, respectively, and a marginal change in the Pd(Cu)/In composition does not significantly influence the ordering temperatures. (orig.).

  11. Development of accident tolerant FeCrAl-ODS steels utilizing Ce-oxide particles dispersion

    Science.gov (United States)

    Shibata, Hiroki; Ukai, Shigeharu; Oono, Naoko H.; Sakamoto, Kan; Hirai, Mutsumi

    2018-04-01

    FeCrAl-ODS ferritic steels with Ce-oxide dispersion instead of Y-oxide were produced for the accident tolerant fuel cladding of the light water reactor. Excess oxygen (Ex.O) was added to improve the mechanical property. The tensile strength at Ex.O = 0 is around 200 MPa at 700 °C, mainly owing to dispersed Ce2O3 particles in less than 10 nm size. The formation of the fine Ce2O3 particles is dominated by a coherent interface with ferritic matrix. With increasing Ex.O, an increased of number density of coarser Ce-Al type oxide particles over 10 nm size is responsible for the improvement of the tensile strength. Change of the type of oxide particle, CeO2, Ce2O3, CeAlO3, Al2O3, in FeCrAl-ODS steel was thermodynamically analyzed as a parameter of Ex.O.

  12. Mirror machine reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1976-01-01

    Recent mirror reactor conceptual design studies are described. Considered in detail is the design of ''standard'' Yin-Yang fusion power reactors with classical and enhanced confinement. It is shown that to be economically competitive with estimates for other future energy sources, mirror reactors require a considerable increase in Q, or major design simplifications, or preferably both. These improvements may require a departure from the ''standard'' configuration. Two attractive possibilities, both of which would use much of the same physics and technology as the ''standard'' mirror, are the field reversed mirror and the end-stoppered mirror

  13. Effluent standards

    Energy Technology Data Exchange (ETDEWEB)

    Geisler, G C [Pennsylvania State University (United States)

    1974-07-01

    At the conference there was a considerable interest in research reactor standards and effluent standards in particular. On the program, this is demonstrated by the panel discussion on effluents, the paper on argon 41 measured by Sims, and the summary paper by Ringle, et al. on the activities of ANS research reactor standards committee (ANS-15). As a result, a meeting was organized to discuss the proposed ANS standard on research reactor effluents (15.9). This was held on Tuesday evening, was attended by members of the ANS-15 committee who were present at the conference, participants in the panel discussion on the subject, and others interested. Out of this meeting came a number of excellent suggestions for changes which will increase the utility of the standard, and a strong recommendation that the effluent standard (15.9) be combined with the effluent monitoring standard. It is expected that these suggestions and recommendations will be incorporated and a revised draft issued for comment early this summer. (author)

  14. EPR study of concentration dependence in Ce, Ce : La and Ce:Y doped SrF2

    NARCIS (Netherlands)

    Dankert, O.; Vainchtein, David; Datema, H.C.; den Hartog, Hendrik

    1995-01-01

    Experimental results of an EPR-study of the concentration dependence of the doubly integrated intensity and linewidth of the signals associated with tetragonal Ce3+-F--dipoles in Sr1-xCexF2+x, Sr-1-0.005-x Ce0.005LaxF2+0.005+x and Sr-1-0.005-x Ce0.005YxF2+0.005+x are presented. Both show a nonlinear

  15. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  16. Comparison of Efficiencies and Mechanisms of Catalytic Ozonation of Recalcitrant Petroleum Refinery Wastewater by Ce, Mg, and Ce-Mg Oxides Loaded Al2O3

    Directory of Open Access Journals (Sweden)

    Chunmao Chen

    2017-02-01

    Full Text Available The use of catalytic ozonation processes (COPs for the advanced treatment of recalcitrant petroleum refinery wastewater (RPRW is rapidly expanding. In this study, magnesium (Mg, cerium (Ce, and Mg-Ce oxide-loaded alumina (Al2O3 were developed as cost efficient catalysts for ozonation treatment of RPRW, having performance metrics that meet new discharge standards. Interactions between the metal oxides and the Al2O3 support influence the catalytic properties, as well as the efficiency and mechanism. Mg-Ce/Al2O3 (Mg-Ce/Al2O3-COP reduced the chemical oxygen demand by 4.7%, 4.1%, 6.0%, and 17.5% relative to Mg/Al2O3-COP, Ce/Al2O3-COP, Al2O3-COP, and single ozonation, respectively. The loaded composite metal oxides significantly increased the hydroxyl radical-mediated oxidation. Surface hydroxyl groups (–OHs are the dominant catalytic active sites on Al2O3. These active surface –OHs along with the deposited metal oxides (Mg2+ and/or Ce4+ increased the catalytic activity. The Mg-Ce/Al2O3 catalyst can be economically produced, has high efficiency, and is stable under acidic and alkaline conditions.

  17. Some considerations about standardization

    Energy Technology Data Exchange (ETDEWEB)

    Dewez, Ph L; Fanjas, Y R [C.E.R.C.A., Romans (France)

    1985-07-01

    Complete standardization of research reactor fuel is not possible. However the transition from HEU to LEU should be an opportunity for a double effort towards standardization and optimization in order to reduce cost. (author)

  18. Some considerations about standardization

    International Nuclear Information System (INIS)

    Dewez, Ph.L.; Fanjas, Y.R.

    1985-01-01

    Complete standardization of research reactor fuel is not possible. However the transition from HEU to LEU should be an opportunity for a double effort towards standardization and optimization in order to reduce cost. (author)

  19. Information System through ANIS at CeSAM

    Science.gov (United States)

    Moreau, C.; Agneray, F.; Gimenez, S.

    2015-09-01

    ANIS (AstroNomical Information System) is a web generic tool developed at CeSAM to facilitate and standardize the implementation of astronomical data of various kinds through private and/or public dedicated Information Systems. The architecture of ANIS is composed of a database server which contains the project data, a web user interface template which provides high level services (search, extract and display imaging and spectroscopic data using a combination of criteria, an object list, a sql query module or a cone search interfaces), a framework composed of several packages, and a metadata database managed by a web administration entity. The process to implement a new ANIS instance at CeSAM is easy and fast : the scientific project has to submit data or a data secure access, the CeSAM team installs the new instance (web interface template and the metadata database), and the project administrator can configure the instance with the web ANIS-administration entity. Currently, the CeSAM offers through ANIS a web access to VO compliant Information Systems for different projects (HeDaM, HST-COSMOS, CFHTLS-ZPhots, ExoDAT,...).

  20. Study of CeI3 evaporation in the presence of group 13 metal-iodides

    International Nuclear Information System (INIS)

    Curry, J. J.; Lapatovich, W. P.; Henins, A.; Hardis, J. E.; Estupiñán, E. G.; Gibbs, J. M.; Shastri, S. D.

    2014-01-01

    The influences of GaI 3 , InI, and TlI on the evaporation characteristics of CeI 3 have been studied over the temperature range 900 K to 1400 K using x-ray induced fluorescence. The total vapor densities, summed over all atomic and molecular species, of Ce, I, In, and Tl were obtained. Measurements of Ce were limited to temperatures above 1033 K, the melting temperature of CeI 3 . This is the highest temperature range for which measurements of the vapor pressure of CeI 3 have been made. The vapor pressure of the CeI 3 monomer above the pure CeI 3 salt for temperatures exceeding its melting point can be approximated by log 10 p/Pa=11.24(±0.03)−10,690(±40) (T/K) −1 where the numbers in parentheses are standard uncertainties. InI and TlI were shown to modestly enhance the presence of Ce in the vapor phase, up to a factor of 5. GaI 3 produced no enhancement in this temperature range. Numerical simulations of the thermochemical equilibrium suggest the importance of both liquid-phase and vapor-phase complexes. Significant improvement to the method of absolute calibration is discussed

  1. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    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.

  2. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    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)

  3. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  4. Ce-Fe-O mixed oxide as oxygen carrier for the direct partial oxidation of methane to syngas

    Institute of Scientific and Technical Information of China (English)

    魏永刚; 王华; 李孔斋

    2010-01-01

    The Ce-Fe-O mixed oxide with a ratio of Ce/Fe=7:3, which was prepared by coprecipitation method and employed as oxygen carrier, for direct partial oxidation of methane to syngas in the absence of gaseous oxygen was explored. The mixed oxide was characterized by X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS) and scanning electron microscopy (SEM), and the catalytic performances were studied in a fixed-bed quartz reactor and a thermogravimetric reactor, respectively. Approximately 99.4% H2 se...

  5. Coherence Kondo gap in CeNiSn and CeRhSb

    International Nuclear Information System (INIS)

    Takabatake, T.; Nakamoto, G.; Tanaka, H.; Bando, Y.; Fujii, H.; Nishigori, S.; Goshima, H.; Suzuki, T.; Fujita, T.; Oguro, I.; Hiraoka, T.; Malik, S.K.

    1994-01-01

    CeNiSn and CeRhSb are Kondo-lattice compounds showing the behavior of a small-gap semiconductor at temperatures below 7 K. We review and discuss the magnetic, transport and specific-heat measurements performed on single crystals of CeNiSn and polycrystals of CeRhSb. Prerequisites for gap formation are deduced from the effects of substitution and application of a magnetic field and pressure on the gapped state. ((orig.))

  6. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  7. The role of Ce(III) in BZ oscillating reactions

    Science.gov (United States)

    Nogueira, Paulo A.; Varela, Hamilton; Faria, Roberto B.

    2012-03-01

    Herein we present results on the oscillatory dynamics in the bromate-oxalic acid-acetone-Ce(III)/Ce(IV) system in batch and also in a CSTR. We show that Ce(III) is the necessary reactant to allow the emergence of oscillations. In batch, oscillations occur with Ce(III) and also with Ce(IV), but no induction period is observed with Ce(III). In a CSTR, no oscillations were found using a freshly prepared Ce(IV), but only when the cerium-containing solution was aged, allowing partial conversion of Ce(IV) to Ce(III) by reaction with acetone.

  8. Adduct formation in Ce(IV) thenolytrifluoroacetonate

    International Nuclear Information System (INIS)

    Anufrieva, S.I.; Polyakova, G.V.; Snezhko, N.I.; Pechurova, N.I.; Martynenko, L.I.; Spitsyn, V.I.

    1982-01-01

    The literature contains no information on adduct formation in Ce(IV) β-diketonates with additional ligands. Since tetrakis-β-diketonates of Ce(IV) have four six-membered chelate rings, we can suppose that the introduction of an additional monodentate or bidentate ligand into the coordination sphere of Ce(IV) β-diketonates would lead to an increase in the coordination number (CN) of the Ce(IV) to nine or ten. The possibility of realization of such a high CN for Ce(IV) has not been proved; a study of adduct formation by Ce(IV) tetrakis-β-diketonates is thus of theoretical interest. Such an investigation might also be of practical interest, because the introduction of an additional ligand into the coordination sphere of a rare-earth β-diketonate usually increases the solubility of the β-diketonate in nonpolar solvents and increases the volatility of the compound; such a modification of the properties is important for various practical purposes. The aim of our work was to study the possibility of separating solid adducts of Ce(IV) tetrakis-thenoyltrifluoroacetonate with certain oxygen-containing and nitrogen-containing donor monodentate and bidentate ligands, and also to investigate their properties. As the β-diketone we used thenoyltrifluoroacetone (HTTFA), since in a parallel investigation it was found that Ce(TTFA) 4 has a high oxidation-reduction stability

  9. Radon gamma-ray spectrometry with YAP:Ce scintillator

    CERN Document Server

    Plastino, W; De Notaristefani, F

    2002-01-01

    The detection properties of a YAP:Ce scintillator (YAlO sub 3 :Ce crystal) optically coupled to a Hamamatsu H5784 photomultiplier with standard bialkali photocathode have been analyzed. In particular, the application to radon and radon-daughters gamma-ray spectrometry was investigated. The crystal response has been studied under severe extreme conditions to simulate environments of geophysical interest, particularly those found in geothermal and volcanic areas. Tests in water up to a temperature of 100 deg.C and in acids solutions such as HCl (37%), H sub 2 SO sub 4 (48%) and HNO sub 3 (65%) have been performed. The measurements with standard radon sources provided by the National Institute for Metrology of Ionizing Radiations (ENEA) have emphasized the non-hygroscopic properties of the scintillator and a small dependence of the light yield on temperature and HNO sub 3. The data collected in this first step of our research have pointed out that the YAP:Ce scintillator can allow high response stability for rad...

  10. Identification and Quality Assessment of Chrysanthemum Buds by CE Fingerprinting

    Directory of Open Access Journals (Sweden)

    Xiaoping Xing

    2015-01-01

    Full Text Available A simple and efficient fingerprinting method for chrysanthemum buds was developed with the aim of establishing a quality control protocol based on biochemical makeup. Chrysanthemum bud samples were successively extracted by water and alcohol. The fingerprints of the chrysanthemum buds samples were obtained using capillary electrophoresis and electrochemical detection (CE-ED employing copper and carbon working electrodes to capture all of the chemical information. 10 batches of chrysanthemum buds were collected from different regions and various factories to establish the baseline fingerprint. The experimental data of 10 batches electropherogram buds by CE were analyzed by correlation coefficient and the included angle cosine methods. A standard chrysanthemum bud fingerprint including 24 common peaks was established, 12 from each electrode, which was successfully applied to identify and distinguish between chrysanthemum buds from 2 other chrysanthemum species. These results demonstrate that fingerprint analysis can be used as an important criterion for chrysanthemum buds quality control.

  11. Standard Technical Specifications, Combustion Engineering plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (M) for Combustion Engineering (CE) Plants and documents the positions of the Nuclear Regulatory Commission based on the CE Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS

  12. Standard Technical Specifications, Combustion Engineering plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for Combustion Engineering (CE) Plants and documents the positions of the Nuclear Regulatory Commission based on the CE Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved SM. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3 contains the Bases for Sections 3.4--3.9 of the improved M

  13. 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)

    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

  14. 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.

  15. Unstable magnetic moments in Ce compounds

    International Nuclear Information System (INIS)

    Aarts, J.

    1984-01-01

    The problems which are connected with the appearance or disappearance of local moments in metals are well reflected in the magnetic behaviour of Ce intermetallic compounds. This work describes experiments on two Ce compounds which are typical examples of unstable moment systems. The first of these is CeAl 2 which at low temperatures, shows coexistence of antiferromagnetic order and the Kondo effect. Measurements are presented of the magnetization and the susceptibility in different magnetic field and temperature regions. An analysis of these measurements, using a model for the crystal field effects, shows the agreement between the measurements and the calculations to be reasonably good for CeAl 2 , but this agreement becomes worse upon decreasing Ce concentration. A phenomenological description of the observations is given. The second compound reported on is CeCu 2 Si 2 , the first 'heavy-fermion' superconductor to be investigated. The superconducting state is possibly formed by the quasi-particles of a non-magnetic many body singlet state, and not simply by the (sd) conduction electrons. This being a novel phenomenon, a number of experiments were performed to test this picture and to obtain a detailed description of the behaviour of CeCu 2 Si 2 . Measurements of the Meissner volume, confirmed the superconductivity to be intrinsic. (Auth.)

  16. Chemically abrupt interface between Ce oxide and Fe films

    International Nuclear Information System (INIS)

    Lee, H.G.; Lee, D.; Kim, S.; Kim, S.G.; Hwang, Chanyong

    2005-01-01

    A chemically abrupt Fe/Ce oxide interface can be formed by initial oxidation of an Fe film followed by deposition of Ce metal. Once a Ce oxide layer is formed on top of Fe, it acts a passivation barrier for oxygen diffusion. Further deposition of Ce metal followed by its oxidation preserve the abrupt interface between Ce oxide and Fe films. The Fe and Ce oxidation states have been monitored at each stage using X-ray photoelectron spectroscopy

  17. Pyrazolates advance cerium chemistry: a CeIII/CeIV redox equilibrium with benzoquinone.

    Science.gov (United States)

    Werner, Daniel; Deacon, Glen B; Junk, Peter C; Anwander, Reiner

    2017-05-16

    Two stable cerium(iv) 3,5-dialkylpyrazolate complexes are presented, namely dimeric [Ce(Me 2 pz) 4 ] 2 (Me 2 pz = 3,5-dimethylpyrazolate) and monomeric Ce(tBu 2 pz) 4 (tBu 2 pz = 3,5-di-tert-butylpyrazolate) along with their trivalent counterparts [Ce(Me 2 pz) 3 ] and [Ce(tBu 2 pz) 3 ] 2 . All complexes were obtained from protonolysis reactions employing the silylamide precursors Ce[N(SiHMe 2 ) 2 ] 4 and Ce[N(SiMe 3 ) 2 ] 3 . Treatment of homoleptic Ce IV and Ce III Me 2 pz complexes with 1,4-hydroquinone (H 2 hq) or 1,4-benzoquinone (bq), respectively, ultimately gave the same trimetallic Ce III species via a cerium redox equilibrium. The Ce III complex Ce 3 (Me 2 pz) 5 (pchd) 2 (L) (pchd = 1,4-bis(3,5-dimethylpyrazol-1-yl)cyclohex-2,5-diene-1,4-diolato; L = Me 2 pzH or (thf) 2 ) results from a di-1,4-pyrazolyl attack on pre-coordinated bq. The reduction of bq by [Ce(Me 2 pz) 3 (thf)] 2 , and re-oxidation by the resulting Ce IV species was supported by UV-vis spectroscopic investigations. Comparisons with the redox-innocent complexes [Ln(Me 2 pz) 3 (thf)] 2 (Ln = La and Pr) revealed far less selective reactions with bq, giving hexametallic and octametallic rare-earth metal side products containing 2-Me 2 pz substituted hq ligands.

  18. Role of the C-E owner group in the nuclear industry

    International Nuclear Information System (INIS)

    Gasper, J.K.

    1988-01-01

    The Combustion Engineering (C-E) owner group was formed in 1979 to address the technical and regulatory issues for plants with nuclear steam supply systems supplied by C-E that resulted from the accident at the Three Mile Island plant. The group addressed many of the immediate concerns from the accident including the response of C-E owner-group plants to small-break loss-of-coolant accidents, the use of power-operated relief valves, emergency operating procedure upgrades, as well as transient and accident responses. The group was successful in addressing these technical and regulatory issues with the US Nuclear Regulatory Commission in such a manner that C-E owner-group plants were not required to shut down or operate at reduced power levels. In recent years, the group has increasingly pursued programs for the benefit of its members rather than because of a regulatory commitment. The major current activities of the group are programs to support the design basis reconstruction of our member plants, improvements to emergency procedure guidelines, which are the basis for the emergency operating procedures used in our plants, a charging pump performance improvement program, various activities to reduce the number of unplanned reactor scrams at our plants, modeling of steam generator corrosion mechanisms to improve steam generator performance, and leak-before-break justification of reactor coolant system piping

  19. Operator licensing examiner standards

    International Nuclear Information System (INIS)

    1993-01-01

    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

  20. Use of standard spectra for the short life radionuclides and ratios for long life radionuclides in the wastes of EDF PWR type reactors

    International Nuclear Information System (INIS)

    Lantes, B.; Bienvenu, Ph.

    2001-01-01

    This paper presents the type of declaration of radioactivity in the wastes of PWR type reactors park. Particularly, it insists on the justification of use of spectra for the declaration of short live radionuclides. It tackles the important developments of methods and measures of radiochemical analysis made by the Cea in order to determine the ratios to declare the long life radioisotopes. (N.C.)

  1. Radiation streaming with SAM-CE

    International Nuclear Information System (INIS)

    De Gangi, N.; Cohen, M.O.; Waluschka, E.; Steinberg, H.A.

    1980-01-01

    The SAM-CE Monte Carlo code has been employed to calculate doses, due to neutron streaming, on the operating floor and other locations of the Millstone Unit II Nuclear Power Facility. Calculated results were compared against measured doses

  2. (EC+β+) decay of 130Ce

    International Nuclear Information System (INIS)

    Xu Shuwei; Zhang Tianmei; Xie Yuanxiang; Ma Ruichang; Ge Yuanxiu; Guo Yingxiang; Wang Chunfang; Li Zhankui; Guo Bing; Xing Jianping; Guo Tianrui; Zhu Shaofei; Xu Wang; Du Jinzhou

    1996-01-01

    The nuclide 130 Ce was produced by a ( 16 O, 4n) reaction on an enriched 118 Sn target. Reaction products were transported to a shielded location by using a helium-jet tape transport system. A 22.9 min activity in chemically separated cerium sample was identified as 130 Ce. The (EC+β + ) decay scheme of 130 Ce was proposed for the first time. This scheme includes 108 γ-lines, 107 γ-lines among them being new. More than 13 1 + low-lying states of 130 La are populated in the decay of 130 Ce. Two new isomers with half-life of 77±10 ns and 17±5 ns were observed by means of delayed γ-γ coincidence measurements. (orig.). With 5 figs., 3 tabs

  3. Formation of Broensted acids sites in the reaction of cyclohexanol on NaCeY zeolites

    International Nuclear Information System (INIS)

    Vogt, O.; Nattich, M.; Datka, J.; Gil, B.

    2002-01-01

    This study was undertaken to elucidate why the catalytic activity of NaCeY in cyclohexanol reactions carried out in a pulse reactor increases with the pulse number. We studied therefore the effect of cyclohexanol and also of ethanol and water on catalytic activity NaCeY (of exchange degrees 36 and 72%) in cyclohexanol reactions: isomerization and disproportionation. We also studied the reaction of cyclohexanol and water with NaCeY zeolite by IR spectroscopy. Our results evidenced that new Broensted acid sites were formed by the reaction of cyclohexanol and water. This was shown by IR spectroscopy: the increase of Si-O 1 H-Al band 3638 cm -1 and in increase of ammonium ions band (upon ammonia adsorption). The new sites were formed by hydrolysis of Ce 3+ ions with water introduced in a pulse, or produced by dehydration of cyclohexanol catalyzed by acid sites. Formation of new Broensted acid sites resulted in an increase of catalytic activity of NaCeY in cyclohexane reaction as observed in this study and also in cyclohexanol reactions. (author)

  4. An Update on Improvements to NiCE Support for PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McCaskey, Alexander J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Billings, Jay Jay [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The Department of Energy Office of Nuclear Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has supported the development of the NEAMS Integrated Computational Environment (NiCE), a modeling and simulation workflow environment that provides services and plugins to facilitate tasks such as code execution, model input construction, visualization, and data analysis. This report details the development of workflows for the reactor core neutronics application, PROTEUS. This advanced neutronics application (primarily developed at Argonne National Laboratory) aims to improve nuclear reactor design and analysis by providing an extensible and massively parallel, finite-element solver for current and advanced reactor fuel neutronics modeling. The integration of PROTEUS-specific tools into NiCE is intended to make the advanced capabilities that PROTEUS provides more accessible to the nuclear energy research and development community. This report will detail the work done to improve existing PROTEUS workflow support in NiCE. We will demonstrate and discuss these improvements, including the development of flexible IO services, an improved interface for input generation, and the addition of advanced Fortran development tools natively in the platform.

  5. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

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

  6. 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.

  7. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  8. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  9. Structural and magnetic properties of Ce/Fe and Ce/FeCoV multilayers

    Energy Technology Data Exchange (ETDEWEB)

    Tixier, S; Boeni, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Mannix, D; Stirling, W G [Liverpool Univ. (United Kingdom); Lander, G H

    1997-09-01

    Ce/Fe and Ce/FeCoV multilayers have been grown by magnetron sputtering. The interfaces are well defined and the layers are crystalline down to an individual layer thickness of 20 A. Ce/FeCoV multilayers show sharper interfaces than Ce/Fe but some loss of crystallinity is observed. Hysteresis loops obtained by SQUID show different behaviour of the bulk magnetisation as a function of the layer thickness. Fe moments are found by Moessbauer spectroscopy to be perpendicular to the interfaces for multilayers with small periodicity. (author) 2 figs., 2 refs.

  10. 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

  11. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    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

  12. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    Zarina Masood

    2011-01-01

    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)

  13. The CE marking in the dimension stone sector: difficulties, contradictions, possible solutions

    Science.gov (United States)

    Primavori, Piero

    2017-04-01

    In accordance with the requirements of the CPR 305/11, no stone products (covered by harmonized standards) can be introduced in the EU market, irrespective of their country of origin, unless they are supported with a Declaration of Performance (DoP) and CE certificate (= CE Marking). The CE marking became compulsory for all stone and marble products as early as 2003, under the legal framework of the CPD 89/106/CE, the EU Directive which, on July 1st, 2013, has been officially replaced by the CPR 305/11. The CE Marking of construction products has been described as one of the most significant change being faced by the construction industry for a decade. Nevertheless, after thirteen years from the introduction of the first products standard, serious difficulties for the CE Marking application still exist. The aim of this contribution is to draw the attention on the effective meaningfulness, applicability and reliability of the CE Marking, on the related aspects for the economic operators (manufacturers, authorized representatives, importers, distributors etc.) and, most of all, for the customers. The following topics and issues are dealt with: - Criteria of the mandatory tests; - Criteria for testing procedures (meaningfulness/reliability/frequency of the TT); - Non-applicability of the testing methods in particular circumstances; - Economic aspects for the companies; - Interpretation of the FPC philosophy; - Formulation of the finished products standards; - Traceability criteria of the stone material; - Threshold-values for the acceptance of a stone material; - Guarantees for the manufacturers and for the customers; - Effective precision and reliability of the DoP and related consequences for manufacturers and customers.

  14. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  15. Mechanism of Methane Chemical Looping Combustion with Hematite Promoted with CeO 2

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Duane D.; Siriwardane, Ranjani

    2013-08-15

    Chemical looping combustion (CLC) is a promising technology for fossil fuel combustion that produces sequestration-ready CO{sub 2} stream, reducing the energy penalty of CO{sub 2} separation from flue gases. An effective oxygen carrier for CLC will readily react with the fuel gas and will be reoxidized upon contact with oxygen. This study investigated the development of a CeO{sub 2}-promoted Fe{sub 2}O{sub 3}-hematite oxygen carrier suitable for the methane CLC process. Composition of CeO{sub 2} is between 5 and 25 wt % and is lower than what is generally used for supports in Fe{sub 2}O{sub 3} carrier preparations. The incorporation of CeO{sub 2} to the natural ore hematite strongly modifies the reduction behavior in comparison to that of CeO{sub 2} and hematite alone. Temperature-programmed reaction studies revealed that the addition of even 5 wt % CeO{sub 2} enhances the reaction capacity of the Fe{sub 2}O{sub 3} oxygen carrier by promoting the decomposition and partial oxidation of methane. Fixed-bed reactor data showed that the 5 wt % cerium oxides with 95 wt % iron oxide produce 2 times as much carbon dioxide in comparison to the sum of carbon dioxide produced when the oxides were tested separately. This effect is likely due to the reaction of CeO{sub 2} with methane forming intermediates, which are reactive for extracting oxygen from Fe{sub 2}O{sub 3} at a considerably faster rate than the rate of the direct reaction of Fe{sub 2}O{sub 3} with methane. These studies reveal that 5 wt % CeO{sub 2}/Fe{sub 2}O{sub 3} gives stable conversions over 15 reduction/oxidation cycles. Lab-scale reactor studies (pulsed mode) suggest the methane reacts initially with CeO{sub 2} lattice oxygen to form partial oxidation products (CO + H{sub 2}), which continue to react with oxygen from neighboring Fe{sub 2}O{sub 3}, leading to its complete oxidation to form CO{sub 2}. The reduced cerium oxide promotes the methane decomposition reaction to form C + H{sub 2}, which continue to

  16. Scintillation properties of Ca co-doped L(Y)SO:Ce between 193 K and 373 K for TOF-PET/MRI

    International Nuclear Information System (INIS)

    Weele, David N ter; Schaart, Dennis R; Dorenbos, Pieter

    2014-01-01

    Time-of-flight Positron Emission Tomography (TOF-PET) and TOF-PET/MRI require scintillators with high light yield, short decay time, and short rise time in order to obtain high timing resolution. LSO:Ce and LYSO:Ce are commonly used. Ca co-doped LSO:Ce shows improved scintillation properties. The decay time constant of LSO:Ce,0.2%Ca (~33 ns) is shorter than standard LSO:Ce (~38-40 ns), and it has about 15% higher light yield. We measured scintillation pulse shapes and photoelectron yields of LSO:Ce, LSO:Ce,0.2%Ca, LYSO:Ce, LYSO:Ce,20ppmCa, LYSO:0.11%Ce,0.2%Mg, and LYSO:0.2%Ce,0.2%Ca at temperatures ranging from 193 K to 373 K. To study rise times we built a set-up in which samples are excited by 100 ps (FWHM) x-ray pulses.

  17. Resonant photoemission study of CeRu4Sb12

    International Nuclear Information System (INIS)

    Ishii, Hiroyoshi; Miyahara, Tsuneaki; Takayama, Yasuhiro; Shiozawa, Hidetsugu; Obu, Kenji; Matsuda, Tatsuma D.; Aoki, Yuji; Sugawara, Hitoshi; Sato, Hideyuki

    2005-01-01

    We have measured the Ce 4d-4f and Ce 3d-4f resonant photoemission spectra of CeRu 4 Sb 12 . The Ce 4f spectra show the spectral features corresponding to a weakly hybridized system. The number of 4f electrons is estimated to be ∼1.0

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  19. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  20. CER. Research reactors in France

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2012-01-01

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  1. Mirror reactor studies

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Bender, D.J.

    1977-01-01

    Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

  2. Macrodynamic study and catalytic reduction of NO by ammonia under mild conditions over Pt-La-Ce-O/Al2O3 catalysts

    International Nuclear Information System (INIS)

    Wang, Yanhui; Zhu, Jingli; Ma, Runyu

    2007-01-01

    Catalytic reduction of NO using ammonia upon series prepared catalysts under 423-573 K in a fixed bed reactor was investigated. Results showed that the performance of supported platinum catalyst could be improved by addition of La and Ce to it. Experimental studies indicated that the suitable molar ratio of Pt:La:Ce would be 1.0:3.78:3.56, Pt-La-Ce (c). Results also found Pt-La-Ce (c) catalyst had good stability and tolerance to certain amounts of sulfur compounds under the used experimental conditions. Characterization for the fresh and used catalysts showed the Pt-La-Ce (c) catalyst had a stable structure. In addition, based on experimental data and using a nonlinear regression algorithm method, an empirical macrodynamic equation was obtained in this study

  3. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  4. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  5. Influence of dopant concentration on spectroscopic properties of Sr2CeO4:Yb nanocrystals

    Science.gov (United States)

    Stefanski, M.; Kędziorski, A.; Hreniak, D.; Strek, W.

    2017-12-01

    Optical properties of Sr2CeO4:Yb nanocrystals synthesized via Pechini's method are reported. The samples were characterized by X-ray diffraction data measurements. The unit cell parameters were determined using Rietveld refinement. It was found that they decreased with increasing amount of Yb ions. The absorption, excitation, emission spectra and luminescence decay profiles of the Sr2CeO4:Yb nanocrystals were investigated. It was observed that optical properties were strongly dependent on Yb concentration. It was found that Yb3+-O2- charge transfer transitions have great influence on the absorption spectra. It can be seen in the emission spectra that in addition to standard bands/lines corresponding to Ce-O metal-to-ligand charge transfer of Sr2CeO4 and f-f transitions of Yb3+, there is emission band centered at 744 nm. Its intensity depends on the concentration of the dopant. Recorded decay times become shorter with increasing dopant concentration due to the Yb3+ concentration quenching. Excitation spectra indicate the energy transfer from Ce-O charge transfer states to Yb3+2F5/2 state. The issue of appearance of down-conversion process in Sr2CeO4:Yb nanocrystals is considered.

  6. Promoting effect of CeO 2 on cyclohexanol conversion over CeO 2

    Indian Academy of Sciences (India)

    Abstract. CeO2-ZnO materials were prepared by amorphous citrate process and characterized by TGA, XRD, UV-DRS and surface area measurements. TGA showed that the citrate precursors decompose in the range 350-550°C producing CeO2-containing catalytic materials. XRD and DRS results indicated the formation of ...

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  8. Fabrication of Nano-CeO2 and Application of Nano-CeO2 in Fe Matrix Composites

    International Nuclear Information System (INIS)

    Tiebao, W.; Chunxiang, C.; Xiaodong, W.; Guobin, L.

    2010-01-01

    It is expatiated that nano-CeO2 is fabricated by the direct sedimentation method. The components and particles diameter of nano-CeO2 powders are analyzed by XRD and SEM . The thermodynamic analysis and acting mechanism of nano-CeO2 with Al in Fe matrix composites are researched, which shows that the reaction is generated between CeO2 and Al in the composite, that is, 3CeO2+4Al - 2Al2O3+3[Ce], which obtains Al2O3 and active [Ce] during the sintering process. The active [Ce] can improve the performance of CeO2/Fe matrix composites. The suitable amount of CeO2 is about 0.05% in CeO2/Fe matrix composites. SEM fracture analysis shows that the toughness sockets in nano-CeO2/Fe matrix composites are more than those in no-added nano-CeO2 composites, which can explain that adding nano-CeO2 into Fe matrix composite, the toughness of the composite is improved significantly. Applied nano-CeO2 to Fe matrix diamond saw blades shows that Fe matrix diamond saw blade is sharper and of longer cutting life than that with no-added nano-CeO2.

  9. Photoluminescent properties of Sr2CeO4: Eu3+ and Sr2CeO4: Eu2+ phosphors suitable for near ultraviolet excitation

    International Nuclear Information System (INIS)

    Suresh, K.; Poornachandra Rao, N.V.; Murthy, K.V.R.

    2014-01-01

    Powder phosphors of 1 mol% Eu 3+ - and Eu 2+ -doped strontium cerium oxide (Sr 2 CeO 4 ) were synthesized by standard solid-state reaction method. Eu 3+ - and Eu 2+ -doped Sr 2 CeO 4 phosphors fired at 1100 ℃ for 2 h were analysed by X-ray diffraction (XRD) and photoluminescence (PL) techniques. The XRD patterns confirm that the obtained phosphors are a single phase of Sr 2 CeO 4 composed of orthorhombic structure. Room temperature PL excitation spectrum of air-heated Sr 2 CeO 4 : Eu phosphor has exhibited bands at 260, 280 and 350 nm. Whereas the excitation spectrum of Sr 2 CeO 4 : Eu phosphor heated under reducing (carbon) atmosphere exhibited single broadband range from 260 to 390 nm. The (PL) emission peaks of both the phosphors at 467 (blue), 537 (green) and 616 nm (red) generate white light under 260, 280 and 350 nm excitation wavelengths. The Commission International de l'Eclairage (CIE) colour coordinates conforms that these phosphors emitting white light. The results reveal that these phosphors are multifunctional phosphors which emit white light under these excitations that they could be used as white components for display and lamp devices and as well as possible good light-conversion phosphor LEDs under near-ultraviolet (nUV) chip. (author)

  10. Reactor feedwater control device

    International Nuclear Information System (INIS)

    Koshi, Yuji.

    1993-01-01

    In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

  11. Physico-chemical properties of (U,Ce)O2

    International Nuclear Information System (INIS)

    Yamada, K.; Yamanaka, S.; Katsura, M.

    1998-01-01

    The high-temperature X-ray diffraction analysis of (U,Ce)O 2 with CeO 2 contents ranging from 0 to 20 mol.% CeO 2 was performed to obtain the variation of the linear thermal expansion coefficient with the CeO 2 content. Ultrasonic pulse-echo measurements were also carried out from room temperature to 673 K to estimate the change in the mechanical properties of (U,Ce)O 2 with the CeO 2 content. The variation in the linear thermal expansion coefficient at the low CeO 2 content region is more steep than that expected from the linear thermal expansion coefficient of UO 2 and CeO 2 . The Young's and shear moduli of all (U,Ce)O 2 were found to decrease with rising temperature. This was due to the increase of the bond length accompanied by the thermal expansion. Although the lattice parameter decreased with CeO 2 content, the moduli of (U,Ce)O 2 were found to decrease with increasing CeO 2 content at room temperature. These results show that in the range from 0 to 20 mol.% of CeO 2 , as CeO 2 content increases, the bottom of the potential energy in (U,Ce)O 2 is shallower and broader. (orig.)

  12. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  13. 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.

  14. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    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)

  15. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    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

  16. Nuclear standards

    International Nuclear Information System (INIS)

    Fichtner, N.; Becker, K.; Bashir, M.

    1981-01-01

    This compilation of all nuclear standards available to the authors by mid 1980 represents the third, carefully revised edition of a catalogue which was first published in 1975 as EUR 5362. In this third edition several changes have been made. The title has been condensed. The information has again been carefully up-dated, covering all changes regarding status, withdrawal of old standards, new projects, amendments, revisions, splitting of standards into several parts, combination of several standards into one, etc., as available to the authors by mid 1980. The speed with which information travels varies and requires in many cases rather tedious and cumbersome inquiries. Also, the classification scheme has been revised with the goal of better adjustment to changing situations and priorities. Whenever it turned out to be difficult to attribute a standard to a single subject category, multiple listings in all relevant categories have been made. As in previous editions, within the subcategories the standards are arranged by organization (in Categorie 2.1 by country) alphabetically and in ascending numerical order. It covers all relevant areas of power reactors, the fuel cycle, radiation protection, etc., from the basic laws and governmental regulations, regulatory guides, etc., all the way to voluntary industrial standards and codes of pratice. (orig./HP)

  17. Study of CeI{sub 3} evaporation in the presence of group 13 metal-iodides

    Energy Technology Data Exchange (ETDEWEB)

    Curry, J. J., E-mail: jjcurry@nist.gov; Lapatovich, W. P.; Henins, A.; Hardis, J. E. [National Institute of Standards and Technology, 100 Bureau Drive, Gaithersburg, Maryland 20899 (United States); Estupiñán, E. G.; Gibbs, J. M. [OSRAM SYLVANIA Inc., 71 Cherry Hill Drive, Beverly, Massachusetts 01915 (United States); Shastri, S. D. [Advanced Photon Source, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, Illinois 60439 (United States)

    2014-01-21

    The influences of GaI{sub 3}, InI, and TlI on the evaporation characteristics of CeI{sub 3} have been studied over the temperature range 900 K to 1400 K using x-ray induced fluorescence. The total vapor densities, summed over all atomic and molecular species, of Ce, I, In, and Tl were obtained. Measurements of Ce were limited to temperatures above 1033 K, the melting temperature of CeI{sub 3}. This is the highest temperature range for which measurements of the vapor pressure of CeI{sub 3} have been made. The vapor pressure of the CeI{sub 3} monomer above the pure CeI{sub 3} salt for temperatures exceeding its melting point can be approximated by log{sub 10}p/Pa=11.24(±0.03)−10,690(±40) (T/K){sup −1} where the numbers in parentheses are standard uncertainties. InI and TlI were shown to modestly enhance the presence of Ce in the vapor phase, up to a factor of 5. GaI{sub 3} produced no enhancement in this temperature range. Numerical simulations of the thermochemical equilibrium suggest the importance of both liquid-phase and vapor-phase complexes. Significant improvement to the method of absolute calibration is discussed.

  18. Bending properties of Ce-TZP/A nanocomposite clasps for removable partial dentures.

    Science.gov (United States)

    Urano, Shinjiro; Hotta, Yasuhiro; Miyazaki, Takashi; Baba, Kazuyoshi

    2015-01-01

    Ceria-stabilized zirconia/alumina nanocomposite (Ce-TZP/A) has excellent fracture toughness and bending strength that could be useful for partial denture framework application. The aim of this study was to investigate the effects of three-dimensional (3D) geometry on the bending and fatigue properties of a model simulation of Ce-TZP/A clasps. Half oval-shaped Ce-TZP/A rods were prepared in six 3D designs. Specimens were either of standard (width divided by thickness: 2.0/1.0 mm) or flat type (2.5/0.8 mm) cross-sectional areas with taper ratios of 1.0, 0.8, or 0.6. As a comparison, cobalt-chromium (Co-Cr) alloy rods of the same shape as the Ce-TZP/A standard shape rod were prepared. All specimens were subjected to the cantilever test and loaded until fracture. They were also cyclically loaded 106 times with various constant displacements, and the maximum displacement prior to fracture was determined for each specimen. Three-dimensional finite element analysis (3D FEA), simulating the cantilever test, was performed to determine the stress distribution during loading. Specimens with the standard cross-sectional shape exhibited higher rigidity and higher fracture loads than the flat specimens by the cantilever test. In addition, lower taper ratios were consistently associated with larger displacements at fracture. Fatigue tests revealed that the maximum displacement prior to fracture of Ce-TZP/A specimens was comparable to that of Co-Cr alloy specimens. The 3D FEA showed that specimens with a taper ratio of 0.6 had the least stress concentration. Ce-TZP/A clasp specimens with a standard cross-sectional shape and a 0.6 taper ratio exhibited the best bending properties among those tested.

  19. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    International Nuclear Information System (INIS)

    Guigon, B.; Vacelet, H.; Dornbusch, D.

    2000-01-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

  20. Materials for passively safe reactors

    International Nuclear Information System (INIS)

    Simnad, T.

    1993-01-01

    Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

  1. The Fermi surface of CeSb

    International Nuclear Information System (INIS)

    Crabtree, G.W.; Aoki, H.; Joss, W.; Hulliger, F.

    1987-01-01

    This paper uses accurate Fermi surface measurements as a test of hybridization models in CeSb. Detailed measurements of the Fermi surface geometry and effective masses are presented which show a number of unusual properties associated with the magnetic structure and anisotropy. Measurements are compared with predictions of a band structure in which the f-electron is assumed to be local, interacting with the conduction electrons only through anisotropic Coulomb and exchange interactions. This model reproduces all the unusual features observed in the measurements and suggests that hybridization is not essential to describing the electronic properties of CeSb

  2. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  3. The New Sun-Sky-Lunar Cimel CE318-T Multiband Photometer - A Comprehensive Performance Evaluation

    Science.gov (United States)

    Barreto, Africa; Cuevas, Emilio; Granados-Munoz, Maria-Jose; Alados-Arboledas, Lucas; Romero, Pedro M.; Grobner, Julian; Kouremeti, Natalia; Almansa, Antonio F.; Stone, Tom; Toledano, Carlos; hide

    2016-01-01

    This paper presents the new photometer CE318-T, able to perform daytime and night-time photometric measurements using the sun and the moon as light source. Therefore,this new device permits a complete cycle of diurnal aerosol and water vapour measurements valuable to enhance atmospheric monitoring to be extracted. In this study wehave found significantly higher precision of triplets when comparing the CE318-T master instrument and the Cimel AErosol RObotic NET work (AERONET) master (CE318-AERONET) triplets as a result of the new CE318-T tracking system. Regarding the instrument calibration, two new methodologies to transfer the calibration from a reference instrument using only daytime measurements (Sun Ratio and Sun-Moon gain factor techniques) are presented and discussed. These methods allow the reduction of the previous complexities inherent to nocturnal calibration. A quantitative estimation of CE318-T AOD uncertainty by means of error propagation theory during daytime revealed AOD uncertainties (u(sup D)(sub AOD)) for Langley-calibrated instruments similar to the expected values for other reference instruments (0.002-0.009). We have also found u(sup D)(sub AOD) values similar to the values reported in sun photometry for field instruments (approximately 0.015). In the case of the night-time period, the CE318-T-estimated standard combined uncertainty (u(sup N)(sub AOD)) is dependent not only on the calibration technique but also on illumination conditions and the instrumental noise. These values range from 0.011-0.018 for Lunar Langley-calibrated instruments to 0.012-0.021 for instruments calibrated using the Sun Ratio technique. In the case of moon-calibrated instruments using the Sun-Moon gain factor method and sun calibrated using the Langley technique, we found u(sup N)(sub AOD) ranging from 0.016 to 0.017 (up to 0.019 in 440 nm channel), not dependent on any lunar irradiance model. A subsequent performance evaluation including CE318-T and collocated

  4. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

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

  5. CeO2-stabilized tetragonal ZrO2 polycrystals (Ce-TZP ceramics)

    International Nuclear Information System (INIS)

    Andrade Nono, M.C. de.

    1990-12-01

    This work presents the development and the characterization of CeO 2 -stabilized tetragonal ZrO 2 polycrystals (Ce-TZP ceramics), since it is considered candidate material for applications as structural high performance ceramics. Sintered ceramics were fabricated from mixtures of powders containing different CeO 2 content prepared by conventional and nonconventional techniques. These powders and their resultant sintered ceramics were specified by chemical and physical characterization, compactation state and mechanical properties. The chemical characteristics were determined by chemical analysis and the physical characteristics were evaluated by phase content, particle and agglomerate size and aspect, and powder porosity. (author)

  6. Production of 139Ce by the 139La(p,n)139Ce reaction

    International Nuclear Information System (INIS)

    Ishioka, Noriko S.; Sekine, Toshiaki; Izumo, Mishiroku; Hashimoto, Kazuyuki; Kobayashi, Katsutoshi; Matsuoka, Hiromitsu

    2002-01-01

    To produce a carrier-free 139 Ce to be used as an efficiency-calibration source for Ge detectors, a target-preparation method and a chemical separation method were studied. It was found that commercially available powders of lanthanum-oxide and lanthanum metal are applicable to a target material in the nuclear reaction 139 La(p,n) 139 Ce. In the separation of 139 Ce from an irradiated lanthanum target, a solvent-extraction method and an ion-exchange method gave final products in good chemical purity. (author)

  7. Fabrication and characterization of CeO{sub 2} pellets for simulation of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    García-Ostos, C.; Rodríguez-Ortiz, J.A. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Arévalo, C., E-mail: carevalo@us.es [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Cobos, J. [CIEMAT, Avenida Complutense, 40, Madrid (Spain); Gotor, F.J. [Materials Science Institute of Seville (CSIC-US), Av. Américo Vespucio, 49, 41092 Seville (Spain); Torres, Y. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain)

    2016-03-15

    Highlights: • CeO{sub 2} is presented as a surrogate material for UO{sub 2} to study nuclear fuel. • Powder-metallurgy methods are applied to fabricate CeO{sub 2} pellets with controlled porosity. • An optimization of the fabrication parameters is established. • Microstructural and tribo-mechanical characterizations are performed. • Properties are compared to those of the nuclear fuel. - Abstract: Cerium Oxide, CeO{sub 2}, has been shown as a surrogate material to understand irradiated Mixed Oxide (MOX) based matrix fuel for nuclear power plants due to its similar structure, chemical and mechanical properties. In this work, CeO{sub 2} pellets with controlled porosity have been developed through conventional powder-metallurgy process. Influence of the main processing parameters (binder content, compaction pressure, sintering temperature and sintering time) on porosity and volumetric contraction values has been studied. Microstructure and physical properties of sintered compacts have also been characterized through several techniques. Mechanical properties such as dynamic Young's modulus, hardness and fracture toughness have been determined and connected to powder-metallurgy parameters. Simulation of nuclear fuel after reactor utilization with radial gradient porosity is proposed.

  8. Enhancement of photocatalytic properties of TiO2 nanoparticles doped with CeO2 and supported on SiO2 for phenol degradation

    International Nuclear Information System (INIS)

    Hao, Chunjing; Li, Jing; Zhang, Zailei; Ji, Yongjun; Zhan, Hanhui; Xiao, Fangxing; Wang, Dan; Liu, Bin; Su, Fabing

    2015-01-01

    Highlights: • CeO 2 -TiO 2 /SiO 2 composites were prepared via a facile co-precipitation method. • Introduction of SiO 2 support increases the dispersion of CeO 2 -TiO 2 . • CeO 2 -TiO 2 /SiO 2 exhibits an enhanced photocatalytic efficiency for phenol degradation. • Ce 3+ /Ce 4+ pair coexisting in CeO 2 improves electron–hole pairs separation efficiency. - Abstract: A series of CeO 2 -TiO 2 and CeO 2 -TiO 2 /SiO 2 composites were prepared with TiCl 4 and Ce (NO 3 ) 3 ·6H 2 O as precursors via a facile co-precipitation method. The obtained samples were characterized by various techniques such as X-ray diffraction (XRD), nitrogen adsorption (N 2 -BET), Fourier transformation infrared spectrum (FT-IR), scanning electron microscopy (SEM), transmission electron microscopy (TEM), and UV–Vis spectroscopy measurements. The results indicated that TiO 2 doped with CeO 2 and supported on SiO 2 could reduce the crystallite size, inhibit the phase transformation, enhance the thermal stability, and effectively extend the spectral response from UV to visible range. When applied to the phenol photodegradation on a homemade batch reactor with an external cooling jacket, the CeO 2 -TiO 2 /SiO 2 catalysts exhibited significantly enhanced photodegradation efficiency in comparison with commercial Degussa P25 and CeO 2 -TiO 2 . The unique catalytic properties of CeO 2 -TiO 2 /SiO 2 were ascribed to improved electron–hole pairs separation efficiency and formation of more reactive oxygen species owing to the presence of Ce 3+ /Ce 4+ , as well as high dispersion of active component of CeO 2 -TiO 2 as a result of the introduction of SiO 2 support. Furthermore, the catalysts can be easily recovered from the reaction solution by centrifugation and reused for four cycles without significant loss of activity

  9. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1993-11-01

    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

  10. Microneedle Array Interface to CE on Chip

    NARCIS (Netherlands)

    Lüttge, Regina; Gardeniers, Johannes G.E.; Vrouwe, E.X.; van den Berg, Albert; Northrup, M.A.; Jensen, K.F; Harrison, D.J.

    2003-01-01

    This paper presents a microneedle array sampler interfaced to a capillary electrophoresis (CE) glass chip with integrated conductivity detection electrodes. A solution of alkali ions was electrokinetically loaded through the microneedles onto the chip and separation was demonstrated compared to a

  11. Fergusonite-type CeNbO{sub 4+δ}: Single crystal growth, symmetry revision and conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Bayliss, Ryan D. [Department of Materials, Imperial College London, Prince Consort Road, London, SW7 2BP (United Kingdom); Pramana, Stevin S.; An, Tao; Wei, Fengxia; Kloc, Christian L. [School of Materials Science and Engineering, 50 Nanyang Avenue, Nanyang Technological University, 639798 (Singapore); White, Andrew J.P. [Chemical Crystallography Laboratory, Department of Chemistry, Imperial College London, Exhibition Road, London, SW7 2AZ (United Kingdom); Skinner, Stephen J. [Department of Materials, Imperial College London, Prince Consort Road, London, SW7 2BP (United Kingdom); White, Timothy J. [School of Materials Science and Engineering, 50 Nanyang Avenue, Nanyang Technological University, 639798 (Singapore); Baikie, Tom, E-mail: tbaikie@ntu.edu.sg [School of Materials Science and Engineering, 50 Nanyang Avenue, Nanyang Technological University, 639798 (Singapore)

    2013-08-15

    Large fergusonite-type (ABO{sub 4}, A=Ce, B=Nb) oxide crystals, a prototype electrolyte composition for solid oxide fuel cells (SOFC), were prepared for the first time in a floating zone mirror furnace under air or argon atmospheres. While CeNbO{sub 4} grown in air contained CeNbO{sub 4.08} as a minor impurity that compromised structural analysis, the argon atmosphere yielded a single phase crystal of monoclinic CeNbO{sub 4}, as confirmed by selected area electron diffraction, powder and single crystal X-ray diffraction. The structure was determined in the standard space group setting C12/c1 (No. 15), rather than the commonly adopted I12/a1. AC impedance spectroscopy conducted under argon found that stoichiometric CeNbO{sub 4} single crystals showed lower conductivity compared to CeNbO{sub 4+δ} confirming interstitial oxygen can penetrate through fergusonite and is responsible for the higher conductivity associated with these oxides. - Graphical abstract: Large fergusonite-type CeNbO{sub 4} crystals were prepared for the first time in a floating zone mirror furnace. Crystal growth in an argon atmosphere yielded a single phase monoclinic CeNbO4, as confirmed by selected area electron diffraction, powder and single crystal X-ray diffraction. The structure was determined in the standard space group setting C12/c1 (No. 15), rather than the commonly adopted I12/a1. AC impedance spectroscopy found CeNbO{sub 4} single crystals showed lower conductivity compared to CeNbO{sub 4+δ} confirming interstitial oxygen can penetrate through fergusonite and is responsible for the higher conductivity associated with these oxides. Highlights: • Preparation of single crystals of CeNbO{sub 4} using a floating zone mirror furnace. • Correction to the crystal symmetry of the monoclinic form of CeNbO{sub 4}. • Report the conductivity of a single crystal of CeNbO{sub 4}.

  12. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  13. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

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

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  15. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1987-01-01

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  17. Correlated electronic structure of CeN

    Energy Technology Data Exchange (ETDEWEB)

    Panda, S.K., E-mail: swarup.panda@physics.uu.se [Department of Physics and Astronomy, Uppsala University, P.O. Box 516, SE-751 20 Uppsala (Sweden); Di Marco, I. [Department of Physics and Astronomy, Uppsala University, P.O. Box 516, SE-751 20 Uppsala (Sweden); Delin, A. [Department of Physics and Astronomy, Uppsala University, P.O. Box 516, SE-751 20 Uppsala (Sweden); KTH Royal Institute of Technology, School of Information and Communication Technology, Department of Materials and Nano Physics, Electrum 229, SE-164 40 Kista (Sweden); KTH Royal Institute of Technology, Swedish e-Science Research Center (SeRC), SE-100 44 Stockholm (Sweden); Eriksson, O., E-mail: olle.eriksson@physics.uu.se [Department of Physics and Astronomy, Uppsala University, P.O. Box 516, SE-751 20 Uppsala (Sweden)

    2016-04-15

    Highlights: • The electronic structure of CeN is studied within the GGA+DMFT approach using SPTF and Hubbard I approximation. • 4f spectral functions from SPTF and Hubbard I are coupled to explain the various spectroscopic manifestations of CeN. • The calculated XPS and BIS spectra show good agreement with the corresponding experimental spectra. • The contribution of the various l-states and the importance of cross-sections for the photoemission process are analyzed. - Abstract: We have studied in detail the electronic structure of CeN including spin orbit coupling (SOC) and electron–electron interaction, within the dynamical mean-field theory combined with density-functional theory in generalized gradient approximation (GGA+DMFT). The effective impurity problem has been solved through the spin-polarized T-matrix fluctuation-exchange (SPTF) solver and the Hubbard I approximation (HIA). The calculated l-projected atomic partial densities of states and the converged potential were used to obtain the X-ray-photoemission-spectra (XPS) and Bremstrahlung Isochromat spectra (BIS). Following the spirit of Gunnarsson–Schonhammer model, we have coupled the SPTF and HIA 4f spectral functions to explain the various spectroscopic manifestations of CeN. Our computed spectra in such a coupled scheme explain the experimental data remarkably well, establishing the validity of our theoretical model in analyzing the electronic structure of CeN. The contribution of the various l-states in the total spectra and the importance of cross sections are also analyzed in detail.

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

    International Nuclear Information System (INIS)

    Herrera, V.; Zamora R, L.

    1997-01-01

    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)

  19. Near-term feasibility of nuclear reactors for seawater desalting. Coupling of standard condensing nuclear power stations to low-grade heat multieffect distillation plants

    International Nuclear Information System (INIS)

    Adar, J.; Manor, S.; Schaal, M.

    1977-01-01

    The paper describes the horizontal aluminium tube, multieffect distillation process developed by Israel Desalination Engineering Ltd., which is very suitable for the use of low-grade heat from standard condensing nuclear turbines operating at increased back-pressure. A special flash-chamber constitutes a positive barrier against any possible contamination being carried over by the steam exhausted from the turbine to the desalination plant. Flow sheets, heat and mass balances have been prepared for two standard sizes of NSSS and turbines, two back-pressures, and corresponding desalination plants. Only standard equipment is being used in the steam and electricity-producing plant. The desalination plant consists of 6 to 12 parallel double lines, each of them similar to a large prototype now being designed and which will be coupled to an old fossil-fuel power station. Total energy requirements of the desalination plant represent only 19 to 50% of the total water cost as against 75% for a single-purpose plant. Costs are based on actual bids for the power plant and actual estimates for the desalination prototype. The operation is designed to be flexible so that the power plant can be operated either in conjunction with the desalination plant, or as a single-purpose plant. (author)

  20. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

    The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

  1. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    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) [pt

  2. Some regularities of Ce(3) and Ce(4) stabilization in their compounds with β-diketones

    International Nuclear Information System (INIS)

    Pechurova, N.I.; Martynenko, L.I.; Snezhko, N.I.; Anufrieva, S.I.

    1985-01-01

    Adduct formation of cerium (3) and cerium (4) β-diketonates (acetylacetonate, benzoylacetonate, dibenzoylmethanate and thenoyltrifluoroacetonate) with oxygen- and nitrogen-donor ligands (Q-α, α'-dipyridyl, o-phenanthroline, trioctylphosphine oxide and triphenylphosphine oxide) is studied. The compounds obtained as a results of the reactions are studied by means of IR-spectroscopic, derivatographic and X-ray phase methods. It is concluded that composition and thermodynamic stability of adducts of Ce(3) tris-β-diketonates are determined by correlation of donor properties of the basis and additional ligand and stability of adducts to oxidation - as well as by their solubility. Introduction of the additional ligand to the system Ce(4)-β-diketones even in the presence of air oxygen stabilizes Ce(3) and destabilizes Ce(4)

  3. Magnetic and electronic properties in CeTSi3 and CeTGe3 (T: transition metal)

    International Nuclear Information System (INIS)

    Shimoda, T.; Okuda, Y.; Takeda, Y.; Ida, Y.; Miyauchi, Y.; Kawai, T.; Fujie, T.; Sugitani, I.; Thamizhavel, A.; Matsuda, T.D.; Haga, Y.; Takeuchi, T.; Nakashima, M.; Settai, R.; Onuki, Y.

    2007-01-01

    We investigated the magnetic properties of CeTSi 3 (T: Ru, Os, Co, Rh, Ir, Pd and Pt) and CeTGe 3 (T: Co, Rh and Ir) by measuring their electrical resistivity and magnetic susceptibility. CeRuSi 3 , CeOsSi 3 and CeCoSi 3 do not order magnetically, with a large Kondo temperature of about 200K. The other compounds order antiferromagnetically, and are very similar to each other in their magnetic and electronic properties, which is related to a large crystalline electric field (CEF) splitting energy of the 4f electron, about 500K in CeIrSi 3

  4. Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

    International Nuclear Information System (INIS)

    1993-01-01

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW t (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment

  5. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    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 U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  6. Tuning Ce distribution for high performanced Nd-Ce-Fe-B sintered magnets

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Xiaodong [School of Materials Science and Engineering, Xi’an University of Technology, Xi’an 710048 (China); Key Laboratory of Magnetic Materials and Devices, Ningbo Institute of Material Technology and Engineering, Chinese Academy of Science, Ningbo 315201 (China); Guo, Shuai; Chen, Kan; Chen, Renjie; Lee, Don [Key Laboratory of Magnetic Materials and Devices, Ningbo Institute of Material Technology and Engineering, Chinese Academy of Science, Ningbo 315201 (China); You, Caiyin, E-mail: caiyinyou@xaut.edu.cn [School of Materials Science and Engineering, Xi’an University of Technology, Xi’an 710048 (China); Yan, Aru, E-mail: aruyan@nimte.ac.cn [Key Laboratory of Magnetic Materials and Devices, Ningbo Institute of Material Technology and Engineering, Chinese Academy of Science, Ningbo 315201 (China)

    2016-12-01

    A dual-alloy method was applied to tune the distribution of Ce for enhancing the performance of Nd-Ce-Fe-B sintered magnets with a nominal composition of (Nd{sub 0.75}Ce{sub 0.25}){sub 30.5}Fe{sub bal}Al{sub 0.1}Cu{sub 0.1}B. In comparison to the single alloy of (Nd{sub 0.75}Ce{sub 0.25}){sub 30.5}Fe{sub bal}Al{sub 0.1}Cu{sub 0.1}B, the coercivity was enhanced from 10.3 kOe to 12.1 kOe and the remanence was increased from 13.1 kG to 13.3 kG for the magnets with a dual-alloy method. In addition, the remanence temperature coefficient α and coercivity temperature coefficient β were also slightly improved for the magnet with the dual alloys. The results of microstructure characterizations show the uniform distribution of Ce for the magnet with a single alloy, and the coexistence of the Ce-rich and Ce-lean regions for the magnet with the dual alloys. In combinations with the nucleation of reversal domains and magnetic recoil curves, the property enhancement of magnets with a dual-alloy method was well explained. - Highlights: • Improved magnetic properties were obtained in dual-alloy magnet. • This is due to the tuning of Ce distribution and the change in microstructure. • The magnetic hardening effect can be observed in dual-alloy magnet.

  7. Excited-state lifetimes in neutron-rich Ce isotopes from EXILL and FATIMA

    Energy Technology Data Exchange (ETDEWEB)

    Koseoglou, P.; Pietralla, N.; Stoyanka, I.; Kroell, T. [IKP, TU-Darmstadt, Darmstadt (Germany); Werner, V. [IKP, TU-Darmstadt, Darmstadt (Germany); Yale University (United States); Bernards, C.; Cooper, N. [Yale University (United States); Blanc, A.; Jentschel, M.; Koester, U.; Mutti, P.; Soldner, T.; Urban, W. [ILL Grenoble (France); Bruce, A.M.; Roberts, O.J. [University of Brighton (United Kingdom); Cakirli, R.B. [MPIK Heidelberg (Germany); France, G. de [GANIL Caen (France); Humby, P.; Patel, Z.; Podolyak, Zs.; Regan, P.H.; Wilson, E. [University of Surrey (United Kingdom); Jolie, J.; Regis, J.-M.; Saed-Samii, N.; Wilmsen, D. [KP, University of Cologne (Germany); Paziy, V. [Universidad Complutense (Spain); Simpson, G.S. [PSC Grenoble (France); Ur, C.A. [INFN Legnaro (Italy)

    2016-07-01

    {sup 235}U and {sup 241}Pu fission fragments were measured by a mixed spectrometer consisting of high-resolution Ge and fast LaBr{sub 3}(Ce)-scintillator detectors at the high-flux reactor of the ILL. Prompt γ-ray cascades from the nuclei of interest are selected via Ge-Ge-LaBr{sub 3}-LaBr{sub 3} coincidences. The good energy resolution of the Ge allow precise gates to be set, selecting the cascade, hence, the nucleus of interest. The excellent timing performance of the LaBr{sub 3} detectors in combination with the General Centroid Difference method allows the measurement of lifetimes in the ps range in preparation for the FATIMA experiment at FAIR. The first results on neutron-rich Ce isotopes are presented.

  8. Inhibition of human carboxylesterases hCE1 and hiCE by cholinesterase inhibitors.

    Science.gov (United States)

    Tsurkan, Lyudmila G; Hatfield, M Jason; Edwards, Carol C; Hyatt, Janice L; Potter, Philip M

    2013-03-25

    Carboxylesterases (CEs) are ubiquitously expressed proteins that are responsible for the detoxification of xenobiotics. They tend to be expressed in tissues likely to be exposed to such agents (e.g., lung and gut epithelia, liver) and can hydrolyze numerous agents, including many clinically used drugs. Due to the considerable structural similarity between cholinesterases (ChE) and CEs, we have assessed the ability of a series of ChE inhibitors to modulate the activity of the human liver (hCE1) and the human intestinal CE (hiCE) isoforms. We observed inhibition of hCE1 and hiCE by carbamate-containing small molecules, including those used for the treatment of Alzheimer's disease. For example, rivastigmine resulted in greater than 95% inhibition of hiCE that was irreversible under the conditions used. Hence, the administration of esterified drugs, in combination with these carbamates, may inadvertently result in decreased hydrolysis of the former, thereby limiting their efficacy. Therefore drug:drug interactions should be carefully evaluated in individuals receiving ChE inhibitors. Copyright © 2012 Elsevier Ireland Ltd. All rights reserved.

  9. Reactor neutron activation analysis on reference materials from intercomparison runs

    International Nuclear Information System (INIS)

    Pantelica, A.; Salagean, M.

    2003-01-01

    A review of using the Instrumental Neutron Activation Analysis (INAA) technique in our laboratory to determine major, minor and trace elements in mineral and biological samples from international intercomparison runs organised by IAEA Vienna, IAEA-MEL Monaco, 'pb-anal' Kosice, INCT Warszawa and IPNT Krakow is presented. Neutron irradiation was carried out at WWR-S reactor in Bucharest (short and long irradiation) during 1982-1997 and at TRIGA reactor in Pitesti (long irradiation) during the later period. The following type of materials were analysed: soils, marine sediments, uranium phosphate ore, water sludge, copper flue dust, whey powder, yeast, cereal flour (rye and wheat), marine animal tissue (mussel, garfish and tuna fish), as well as vegetal tissue (seaweed, cabbage, spinach, alfalfa, algae, tea leaves and herbs). The following elements could be, in general, determined: Ag, As, Au, Ba, Br, Ca, Ce, Co, Cr, Cs, Eu, Fe, Hf, Hg, K, La, Lu, Mo, Na, Nd, Ni, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, U, W, Yb and Zn of long-lived radionuclides, as well as Al, Ca, Cl, Cu, Mg, Mn, and Ti of short-lived radionuclides. Data obtained in our laboratory for various matrix samples presented and compared with the intercomparison certified values. The intercomparison exercises offer to the participating laboratories the opportunity to test the accuracy of their analytical methods as well as to acquire valuable Reference Materials/ standards for future analytical applications. (authors)

  10. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Schultz, K.R.; Smith, A.C. Jr.

    1978-01-01

    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

  11. CeO2 nanoparticles induce no changes in phenanthrene toxicity to the soil organisms Porcellionides pruinosus and Folsomia candida.

    Science.gov (United States)

    Tourinho, Paula S; Waalewijn-Kool, Pauline L; Zantkuijl, Irene; Jurkschat, Kerstin; Svendsen, Claus; Soares, Amadeu M V M; Loureiro, Susana; van Gestel, Cornelis A M

    2015-03-01

    Cerium oxide nanoparticles (CeO2 NPs) are used as diesel fuel additives to catalyze oxidation. Phenanthrene is a major component of diesel exhaust particles and one of the most common pollutants in the environment. This study aimed at determining the effect of CeO2 NPs on the toxicity of phenanthrene in Lufa 2.2 standard soil for the isopod Porcellionides pruinosus and the springtail Folsomia candida. Toxicity tests were performed in the presence of CeO2 concentrations of 10, 100 or 1000mg Ce/kg dry soil and compared with results in the absence of CeO2 NPs. CeO2 NPs had no adverse effects on isopod survival and growth or springtail survival and reproduction. For the isopods, LC50s for the effect of phenanthrene ranged from 110 to 143mg/kg dry soil, and EC50s from 17.6 to 31.6mg/kg dry soil. For the springtails, LC50s ranged between 61.5 and 88.3mg/kg dry soil and EC50s from 52.2 to 76.7mg/kg dry soil. From this study it may be concluded that CeO2 NPs have a low toxicity and do not affect toxicity of phenanthrene to isopods and springtails. Copyright © 2014 Elsevier Inc. All rights reserved.

  12. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  13. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  15. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  16. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  17. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    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

  18. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  19. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  20. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  1. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  3. Facile hydrothermal synthesis of CeO 2 nanopebbles

    Indian Academy of Sciences (India)

    Cerium oxide (CeO2) nanopebbles have been synthesized using a facile hydrothermal method. X-ray diffraction pattern (XRD) and transmission electron microscopy analyses confirm the presence of CeO2 nanopebbles. XRD shows the formation of cubic fluorite CeO2 and the average particle size estimated from the ...

  4. Radiation safety assessment and development of environmental radiation monitoring technology; standardization of input parameters for the calculation of annual dose from routine releases from commercial reactor effluents

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I. H.; Cho, D.; Youn, S. H.; Kim, H. S.; Lee, S. J.; Ahn, H. K. [Soonchunhyang University, Ahsan (Korea)

    2002-04-01

    This research is to develop a standard methodology for determining the input parameters that impose a substantial impact on radiation doses of residential individuals in the vicinity of four nuclear power plants in Korea. We have selected critical nuclei, pathways and organs related to the human exposure via simulated estimation with K-DOSE 60 based on the updated ICRP-60 and sensitivity analyses. From the results, we found that 1) the critical nuclides were found to be {sup 3}H, {sup 133}Xe, {sup 60}Co for Kori plants and {sup 14}C, {sup 41}Ar for Wolsong plants. The most critical pathway was 'vegetable intake' for adults and 'milk intake' for infants. However, there was no preference in the effective organs, and 2) sensitivity analyses showed that the chemical composition in a nuclide much more influenced upon the radiation dose than any other input parameters such as food intake, radiation discharge, and transfer/concentration coefficients by more than 102 factor. The effect of transfer/concentration coefficients on the radiation dose was negligible. All input parameters showed highly estimated correlation with the radiation dose, approximated to 1.0, except for food intake in Wolsong power plant (partial correlation coefficient (PCC)=0.877). Consequently, we suggest that a prediction model or scenarios for food intake reflecting the current living trend and a formal publications including details of chemical components in the critical nuclei from each plant are needed. Also, standardized domestic values of the parameters used in the calculation must replace the values of the existed or default-set imported factors via properly designed experiments and/or modelling such as transport of liquid discharge in waters nearby the plants, exposure tests on corps and plants so on. 4 figs., 576 tabs. (Author)

  5. Cerium intermetallics CeTX. Review III

    Energy Technology Data Exchange (ETDEWEB)

    Poettgen, Rainer; Janka, Oliver [Muenster Univ. (Germany). Inst. fuer Anorganische und Analytische Chemie; Chevalier, Bernard [Bordeaux Univ., Pessac (France). Inst. de Chimie de la Matiere Condensee de Bordeaux

    2016-05-01

    The structure-property relationships of CeTX intermetallics with structures other than the ZrNiAl and TiNiSi type are systematically reviewed. These CeTX phases form with electron-poor and electron-rich transition metals (T) and X = Mg, Zn, Cd, Hg, Al, Ga, In, Tl, Si, Ge, Sn, Pb, P, As, Sb, and Bi. The review focusses on the crystal chemistry, the chemical bonding peculiarities, and the magnetic and transport properties. Furthermore {sup 119}Sn Moessbauer spectroscopic data, high-pressure studies, hydrogenation reactions and the formation of solid solutions are reviewed. This paper is the third of a series of four reviews on equiatomic intermetallic cerium compound [Part I: R. Poettgen, B. Chevalier, Z. Naturforsch. 2015, 70b, 289; Part II: R. Poettgen, B. Chevalier, Z. Naturforsch. 2015, 70b, 695].

  6. Evaluation and standardization of neutron activation analysis according to the K{sub 0} method in the RP-10 reactor; Evaluacion y estandarizacion del analisis por activacion neutronica segun el metodo del K{sub 0} en el reactor nuclear RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Montoya R, E

    1995-06-01

    It has been characterized and standardized an irradiation of the RP-10 Research Nuclear Reactor for use of the K{sub 0} 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 K{sub 0} 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.

  7. Experimental measurement of enthalpy increments of Th0.25Ce0.75O2

    International Nuclear Information System (INIS)

    Babu, R.; Balakrishnan, S.; Ananthasivan, K.; Nagarajan, K.

    2013-01-01

    Thorium has been suggested as an alternative fertile material for a nuclear fuel cycle, and an inert matrix for burning plutonium and for waste disposal. The third stage of India's nuclear power programme envisages utilization of thorium and plutonium as a fuel in Advanced Heavy Water Reactor (AHWR) and Accelerator Driven Sub-critical Systems (ADSS). Solid solutions of ThO 2 -PuO 2 are of importance because of coexistence of Th with Pu during the breeding cycle. CeO 2 is used as a PuO 2 analog due to similar ionic radii of cations and similar physico-chemical properties of the oxides. ThO 2 forms a homogeneous solid solution with the cubic fluorite structure when doped with Ce in the entire compositional range. In the development of mixed oxide nuclear fuels, knowledge of thermodynamic properties of thorium oxide and its mixtures has become extremely importance for understanding the fuel behavior during irradiation and for predicting the performance of the fuel under accidental conditions. Thermodynamic functions such as the enthalpy increment and heat capacity of the theria-ceria solid solution have not been measured experimentally. Hence, the enthalpy increments of thoria-ceria solid solutions, Th 0.25 Ce 0.75 O 2 by inverse drop calorimetry in the temperature range 523-1723 K have been measured. The measured enthalpy increments were fitted in to polynomial functions by using the least squares method and the other thermodynamic functions such as heat capacity, entropy and Gibbs energy functions were computed in the temperature range 298-1800 K. The reported thermodynamic functions for Th 0.25 Ce 0.75 O 2 forms the first experimental data and the heat capacity of (Th,Ce)O 2 solid solutions was shown to obey the Neumann-Kopp's rule. (author)

  8. Preparation and characterization of Au/CeO{sub 2}-Al{sub 2}O{sub 3} monoliths

    Energy Technology Data Exchange (ETDEWEB)

    Gawel, Bartlomiej; Lambrechts, Kalle [Ugelstad Laboratory, Department of Chemical Engineering, Norwegian University of Science and Technology (NTNU), N-7491 Trondheim (Norway); Oye, Gisle, E-mail: gisle.oye@chemeng.ntnu.no [Ugelstad Laboratory, Department of Chemical Engineering, Norwegian University of Science and Technology (NTNU), N-7491 Trondheim (Norway)

    2012-05-15

    Highlights: Black-Right-Pointing-Pointer A facile method for preparing Au/CeO{sub 2}-Al{sub 2}O{sub 3} monoliths with hierarchical porosity. Black-Right-Pointing-Pointer Continuous-flow testing of the monoliths in liquid-phase oxidation of glucose. Black-Right-Pointing-Pointer Increased catalytic activity in the presence of cerium oxide (stirred-batch tests). - Abstract: Porous CeO{sub 2}-Al{sub 2}O{sub 3} monoliths with hierarchical pore structure were prepared by mixing boehmite particles with solutions containing different amounts of cerium chloride and aluminum nitrate. The monoliths were functionalized with gold nanoparticles using the incipient wetness method. The resulting materials were characterized by X-ray diffraction, nitrogen sorption, mercury porosimetry, UV-vis spectroscopy and transmission electron microscopy. The catalysts were tested in liquid phase glucose oxidation, comparing continuously stirred batch reactor and continuous-flow fix-bed reactor setups.

  9. Plasma-assisted adsorption of elemental mercury on CeO2/TiO2 at low temperatures

    Science.gov (United States)

    Liu, Lu; Zheng, Chenghang; Gao, Xiang

    2017-11-01

    Mercury is a kind of pollutants contained in flue gas which is hazardous for human beings. In this work, CeO2 was packed in the discharge zone of a plasma reactor to adsorb elemental mercury at low temperatures. Plasma-catalyst reactor can remove Hg0 efficiently with CeO2/TiO2 catalysts packed in the discharge zone. The Hg0 concentration continued to decrease gradually when the plasma was turned on, but not sank rapidly. This tendency was different with other catalysts. The treatment of plasma to CeO2/TiO2 catalysts has a promotion effect on the adsorption of Hg0. Plasma has the effect of changing the surface properties of the catalysts and the changes would restitute if the condition changed. The long-running test demonstrated that this method is an effective way to remove Hg0. The removal efficiency remained at above 99% throughout 12 hours when plasma had been turned on (15kV, 0.5 g packed CeO2/TiO2).

  10. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  11. Coherent neutrino scattering with low temperature bolometers at Chooz reactor complex

    International Nuclear Information System (INIS)

    Billard, J; Gascon, J; Jesus, M De; Carr, R; Formaggio, J A; Heine, S T; Johnston, J; Leder, A; Sibille, V; Winslow, L; Dawson, J; Lasserre, T; Figueroa-Feliciano, E; Palladino, K J; Vivier, M

    2017-01-01

    We present the potential sensitivity of a future recoil detector for a first detection of the process of coherent elastic neutrino nucleus scattering (CE ν NS). We use the Chooz reactor complex in France as our luminous source of reactor neutrinos. Leveraging the ability to cleanly separate the rate correlated with the reactor thermal power against (uncorrelated) backgrounds, we show that a 10 kg cryogenic bolometric array with 100 eV threshold should be able to extract a CE ν NS signal within one year of running. (paper)

  12. 78 FR 75557 - CE FLNG, LLC, CE Pipeline, LLC; Notice of Intent To Prepare an Environmental Impact Statement for...

    Science.gov (United States)

    2013-12-12

    ... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. PF13-11-000] CE FLNG, LLC, CE Pipeline, LLC; Notice of Intent To Prepare an Environmental Impact Statement for the Planned CE FLNG Project, Request for Comments on Environmental Issues, and Notice of Public Scoping Meeting The staff of the Federal Energy Regulatory Commission ...

  13. Acetylene and carbon monoxide oxidation over a Pt/Rh/CeO2/γ-Al2O3 automotive exhaust gas catalyst: kinetic modelling of transient experiments

    NARCIS (Netherlands)

    Harmsen, J.M.A.; Hoebink, J.H.B.J.; Schouten, J.C.

    2001-01-01

    The transient kinetics of acetylene (C2H2) conversion by oxygen over a commercial Pt/Rh/CeO2/¿-Al2O3 three-way catalyst have been modelled. Experiments to validate the model were carried out in a fixed-bed reactor with two separate inlets, enabling alternate feeding of acetylene and oxygen.

  14. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  15. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  16. Itinerant f-electron behavior in Ce and U compounds

    International Nuclear Information System (INIS)

    Crabtree, G.W.

    1985-04-01

    The experimentally observed Fermi surface properties in URh 3 , UIr 3 , UGe 3 , CeSn 3 , CeB 6 , U 3 As 4 , U 3 P 4 , and CeSb are reviewed. For the compounds with no magnetic order, band structure models of the Fermi surface geometry are confirmed and f-ligand hybridization is found to be dominant. For CeB 6 , U 3 As 4 , and U 3 P 4 the experiments show that both local moments and f hybridization are important. In CeSb new data can be explained by a purely local model with no f-hybridization

  17. Thermodynamic stability studies of Ce-Sb compounds with Fe

    Science.gov (United States)

    Xie, Yi; Zhang, Jinsuo; Benson, Michael T.; Mariani, Robert D.

    2018-02-01

    Lanthanide fission products can migrate to the fuel periphery and react with cladding, causing fuel-cladding chemical interaction (FCCI). Adding a fuel additive dopant, such as Sb, can bind lanthanide, such as Ce, into metallic compounds and thus prevent migration. The present study focuses on the thermodynamic stability of Ce-Sb compounds when in contact with the major cladding constituent Fe by conducting diffusion couple tests. Ce-Sb compounds have shown high thermodynamic stability as they did not react with Fe. When Fe-Sb compounds contacted with Ce, Sb was separated out of Fe-Sb compounds and formed the more stable Ce-Sb compounds.

  18. Promoting safety in nuclear installations. The IAEA has established safety standards for nuclear reactors and provides expert review and safety services to assist Member States in their application

    International Nuclear Information System (INIS)

    2002-01-01

    More than 430 nuclear power plants (NPPs) are currently operating in 30 countries around the world. The nuclear share of total electricity production ranges from about 20 percent in the Czech Republic and United States to nearly 78 percent in France and Lithuania. Worldwide, nuclear power generates about 16% of the total electricity. The safety of such nuclear installations is fundamental. Every aspect of a power plant must be closely supervised and scrutinized by national regulatory bodies to ensure safety at every phase. These aspects include design, construction, commissioning, trial operation, commercial operation, repair and maintenance, plant upgrades, radiation doses to workers, radioactive waste management and, ultimately, plant decommissioning. Safety fundamentals comprise defence-in-depth, which means having in place multiple levels of protection. nuclear facilities; regulatory responsibility; communicating with the public; adoption of the international convention on nuclear safety including implementation of IAEA nuclear safety standards. This publication covers topics of designing for safety (including safety concepts, design principles, and human factors); operating safety (including safety culture and advance in operational safety); risk assessment and management

  19. New safety standards of nuclear power station with no requirements of site evaluation. No public dose limit published with possible inappropriateness of reactor site

    International Nuclear Information System (INIS)

    Takitani, Koichi

    2013-01-01

    Nuclear Regulation Authority was preparing new safety standards in order to aim at starting safety reviews of existing nuclear power station in July 2013. This article commented on issues of major accident, which was defined as severely damaged core event. Accumulated dose at the site boundary of the Fukushima Daiichi Nuclear Power Station totaled to about 234 mSv on March just after the accident with rare gas of 500 PBq, iodine 131 of 500 PBq, cesium 134 of 10 PBq and cesium 137 of 10 PBq released to the atmosphere, which was beyond 100 mSv. As measures for preventing containment vessel failure after core severely damaged, filtered venting system was required to be installed for low radiological risk to the public. However filter was not effective to rare gas. Accumulated doses at the site boundary of several nuclear power stations after filtered venting with 100% release of rare gas could be estimated to be 2-37 Sv mostly depending on the site condition, which might be surely greater than 100 mSv. Omitting site evaluation for major accident, which was beyond design basis accident, was great concern. (T. Tanaka)

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

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

    International Nuclear Information System (INIS)

    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

  2. Feed water control device in a reactor

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1984-01-01

    Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

  3. Ce(III)/Ce(IV) in methanesulfonic acid as the positive half cell of a redox flow battery

    International Nuclear Information System (INIS)

    Leung, P.K.; Ponce de Leon, C.; Low, C.T.J.; Walsh, F.C.

    2011-01-01

    The characteristics of the Ce(III)/Ce(IV) redox couple in methanesulfonic acid were studied at a platinum disk electrode (0.125 cm 2 ) over a wide range of electrolyte compositions and temperatures: cerium (III) methanesulfonate (0.1-1.2 mol dm -3 ), methanesulfonic acid (0.1-5.0 mol dm -3 ) and electrolyte temperatures (295-333 K). The cyclic voltammetry experiments indicated that the diffusion coefficient of Ce(III) ions was 0.5 x 10 -6 cm 2 s -1 and that the electrochemical kinetics for the oxidation of Ce(III) and the reduction of Ce(IV) was slow. The reversibility of the redox reaction depended on the electrolyte composition and improved at higher electrolyte temperatures. At higher methanesulfonic acid concentrations, the degree of oxygen evolution decreased by up to 50% when the acid concentration increased from 2 to 5 mol dm -3 . The oxidation of Ce(III) and reduction of Ce(IV) were also investigated during a constant current batch electrolysis in a parallel plate zinc-cerium flow cell with a 3-dimensional platinised titanium mesh electrode. The current efficiencies over 4.5 h of the process Ce(III) to Ce(IV) and 3.3 h electrolysis of the reverse reaction Ce(IV) to Ce(III) were 94.0 and 97.6%, respectively. With a 2-dimensional, planar platinised titanium electrode (9 cm 2 area), the redox reaction of the Ce(III)/Ce(IV) system was under mass-transport control, while the reaction on the 3-dimensional mesh electrode was initially under charge-transfer control but became mass-transport controlled after 2.5-3 h of electrolysis. The effect of the side reactions (hydrogen and oxygen evolution) on the current efficiencies and the conversion of Ce(III) and Ce(IV) are discussed.

  4. Effects of CeO2 nanoparticles on sludge aggregation and the role of extracellular polymeric substances – Explanation based on extended DLVO

    International Nuclear Information System (INIS)

    You, Guoxiang; Hou, Jun; Wang, Peifang; Xu, Yi; Wang, Chao; Miao, Lingzhan; Lv, Bowen; Yang, Yangyang; Luo, Hao

    2016-01-01

    The extended DLVO (XDLVO) theory was applied to elucidate the potential effects of CeO 2 nanoparticles (CeO 2 NPs) on sludge aggregation and the role of extracellular polymeric substances (EPS). In this study, seven different concentrations of CeO 2 NPs were added to activated sludge cultured in sequencing batch reactors (SBRs) and compared with a control test that received no CeO 2 NPs. After exposure to 50 mg/L CeO 2 NPs, a negligible change (p>0.1) occurred in the sludge volume index (SVI), whereas the flocculability and aggregation of the sludge decreased by 18.8% and 11.2%, respectively, resulting in a high effluent turbidity. The XDLVO theory demonstrated that the adverse effects of the CeO 2 NPs on sludge aggregation were due to an enhanced barrier energy. Compared to the van der Waals energies (W A ) and the electric double layer (W R ), the acid-base interaction (W AB ) markedly changed for the various concentrations of CeO 2 NPs. The EPS played a decisive role in the sludge surface characteristics, as the removal of EPS equals to the negative effects induced by 5–10 mg/L CeO 2 NPs on the sludge flocculability and aggregation. The presence of CeO 2 NPs induced negative contributions to the tight boundary EPS (TB-EPS) and core bacteria while positive contributions to the total interaction energy of the loose boundary EPS (LB-EPS). - Highlights: • CeO 2 NPs adversely affected the flocculability and aggregation of the sludge. • The presence of CeO 2 NPs increased the energy barrier and led to a stable suspension. • The removal of EPS equals to the negative effects induced by 5–10 mg/L CeO 2 NPs. • The acid-base interaction was dominate and markedly changed for the CeO 2 NPs. • CeO 2 NPs induced negative contributions to the TB-EPS while positive to the LB-EPS.

  5. Effects of CeO{sub 2} nanoparticles on sludge aggregation and the role of extracellular polymeric substances – Explanation based on extended DLVO

    Energy Technology Data Exchange (ETDEWEB)

    You, Guoxiang [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China); Hou, Jun, E-mail: hhuhjyhj@126.com [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China); Wang, Peifang, E-mail: pfwang2005@hhu.edu.cn [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China); Xu, Yi; Wang, Chao; Miao, Lingzhan; Lv, Bowen; Yang, Yangyang; Luo, Hao [Key Laboratory of Integrated Regulation and Resources Development on Shallow Lakes, Ministry of Education, Hohai University, Nanjing 210098 (China); College of Environment, Hohai University, Nanjing 210098 (China)

    2016-11-15

    The extended DLVO (XDLVO) theory was applied to elucidate the potential effects of CeO{sub 2} nanoparticles (CeO{sub 2} NPs) on sludge aggregation and the role of extracellular polymeric substances (EPS). In this study, seven different concentrations of CeO{sub 2} NPs were added to activated sludge cultured in sequencing batch reactors (SBRs) and compared with a control test that received no CeO{sub 2} NPs. After exposure to 50 mg/L CeO{sub 2} NPs, a negligible change (p>0.1) occurred in the sludge volume index (SVI), whereas the flocculability and aggregation of the sludge decreased by 18.8% and 11.2%, respectively, resulting in a high effluent turbidity. The XDLVO theory demonstrated that the adverse effects of the CeO{sub 2} NPs on sludge aggregation were due to an enhanced barrier energy. Compared to the van der Waals energies (W{sub A}) and the electric double layer (W{sub R}), the acid-base interaction (W{sub AB}) markedly changed for the various concentrations of CeO{sub 2} NPs. The EPS played a decisive role in the sludge surface characteristics, as the removal of EPS equals to the negative effects induced by 5–10 mg/L CeO{sub 2} NPs on the sludge flocculability and aggregation. The presence of CeO{sub 2} NPs induced negative contributions to the tight boundary EPS (TB-EPS) and core bacteria while positive contributions to the total interaction energy of the loose boundary EPS (LB-EPS). - Highlights: • CeO{sub 2} NPs adversely affected the flocculability and aggregation of the sludge. • The presence of CeO{sub 2} NPs increased the energy barrier and led to a stable suspension. • The removal of EPS equals to the negative effects induced by 5–10 mg/L CeO{sub 2} NPs. • The acid-base interaction was dominate and markedly changed for the CeO{sub 2} NPs. • CeO{sub 2} NPs induced negative contributions to the TB-EPS while positive to the LB-EPS.

  6. Restart of R reactor at SRP

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1983-01-01

    Restart of the Savannah River R-Reactor is an alternative to L-Reactor operation for increased production of defense nuclear material. R-Reactor was shut down in 1964 after 11-years operation and has been on standby for 19 years. This report presents a description of R-Reactor operation to serve as a basis for analysis of environmental impacts after restoration to meet current SRP performance standards. R-Reactor operation would differ from L-Reactor operation principally in discharge and recycle of effluent cooling water to Par Pond, rather than direct discharge to the Savannah River by way of Steel Creek. Significant differences in environmental effects could result. A costly renovation program would be required to restore R-Reactor to operability, and the reactor could not contribute to material production before about 1989

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  9. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  10. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  11. 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.

  12. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    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 [fr

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  14. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  15. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  16. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  17. Experience with reactor power cutback system at Palo Verde nuclear generating station

    International Nuclear Information System (INIS)

    Chari, D.R.; Rec, J.R.; Simoni, L.P.; Eimar, R.L.; Sowers, G.W.

    1987-01-01

    Palo Verde Nuclear Generating Station (PVNGS) is a three unit site which illustrates System 80 nuclear steam supply system (NSSS) design. The System 80 NSSS is the Combustion Engineering (C-E) standard design rated at 3817 Mwth. PVNGS Units 1 and 2 achieved commercial operation on February 13, 1986 and September 22, 1986, respectively, while Unit 3 has a forecast date for commercial operation in the third quarter of 1987. The System 80 design incorporates a reactor power cutback system (RPCS) feature which reduces plant trips caused by two common initiating events: loss of load/turbine trip (LOL) and loss of one main feedwater pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety system

  18. Calorimetric dosimetry of reactor radiation

    International Nuclear Information System (INIS)

    Radak, B.; Markovic, V.; Draganic, I.

    1961-01-01

    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 [sr

  19. Stress corrosion cracking (Standard Astm G 30-90) in stainless steel 08X18H10T of swimming-pool that contain nuclear fuel in reactors V.V.E.R.-440

    International Nuclear Information System (INIS)

    Zamora R, L.; Herrera, V.

    1998-01-01

    The standard recommended practice for making and using 'U' bend stress corrosion test specimens; Designation G30-90 has been used as a laboratory tool to study the susceptibility of austenitic stainless steels and the other materials of test of intergranular stress corrosion cracking (IGSCC). The experiment has been development in a similar conditions of the chemical regime, the swimming-pool that containing nuclear fuel in borated water reactors VVER-440 in general this cladding by two films, one of carbon steel (04T26) and other with austenitic stainless steel 08X18HT (similar type 321) stabilized with titanium, the thickness of filler metals was to 4 to 8 mm. The specimens was prepare one plate with this characteristics, the welding was put in the part central with the following measurements of 160x15x5 mm. The specimens strips bent approximately 180 degrees around radius of curvature of R=14.5 mm and ε 1 = 17.2% and maintained in this plastically deformed condition during the test. And then preparing metallographically and exposure in environment of 12 and 40 gr./l of H 3 BO 3 70 Centigrade with or noting contaminants of NaCl. The results showed the initial cracks. (Author)

  20. Transient analysis capabilities at ABB-CE

    International Nuclear Information System (INIS)

    Kling, C.L.

    1992-01-01

    The transient capabilities at ABB-Combustion Engineering (ABB-CE) Nuclear Power are a function of the computer hardware and related network used, the computer software that has evolved over the years, and the commercial technical exchange agreements with other related organizations and customers. ABB-CEA is changing from a mainframe/personal computer network to a distributed workstation/personal computer local area network. The paper discusses computer hardware, mainframe computing, personal computers, mainframe/personal computer networks, workstations, transient analysis computer software, design/operation transient analysis codes, safety (licensed) analysis codes, cooperation with ABB-Atom, and customer support

  1. Synthesis and characterization of Ce doped MFI zeolite

    International Nuclear Information System (INIS)

    Kalita, Banani; Talukdar, Anup K.

    2012-01-01

    Highlights: ► Cerium was incorporated into the tetrahedral position of MFI zeolite structure. ► Unit cell volume increases with an increase of Ce content in the framework of MFI. ► A band at 310 nm in the UV–vis spectra indicates Ce incorporation in MFI structure. ► The mass loss (%) in the region 373–423 K decreases with increase of Ce in MFI. - Abstract: Ce doped MFI (mobil five) zeolites with different Si to (Ce + Al) and different Ce to Al ratios were synthesized by a hydrothermal synthesis method. All the samples were characterized by different techniques such as X-ray diffraction, Fourier transform infrared spectroscopy (FTIR), thermogravimetric analysis (TGA), UV–vis diffuse reflectance spectroscopy (DRS) and scanning electron microscopy (SEM). The results showed that almost pure MFI phase was obtained in all cases with incorporation of cerium metal into the framework. The increase in unit cell parameters observed with an increase in Ce content is indicative of incorporation of Ce into the framework structure of microporous material MFI. Corroboration of the framework incorporation of Ce into the MFI zeolite structure was also obtained from the UV–vis DRS spectra by the presence of an absorption band at 280 nm. TGA and SEM of the samples provide complementary evidence for Ce incorporation into the framework MFI structure.

  2. Synthesis of CeS and interactions with molten metals

    International Nuclear Information System (INIS)

    Krikorian, O.H.; Curtis, P.G.

    1988-01-01

    Hot-pressed and sintered discs of single-phase CeS were tested for interaction with molten aluminium, uranium, and iron to determine the conditions under which reaction first begins and the nature of the reaction. Aluminium begins to react with CeS at ∼ 1190 K, slowly dissolving cerium and forming a thin layer of Ce 3 S 4 at the reaction interface. At 1363 K, aluminium wets and spreads over the CeS surface and dissolves ∼ 01 at% Ce. Ce 3 Al 11 precipitates out in the aluminium phase on cooldown. Uranium does not react with CeS at 1673 K, but at 1873 K it wets and spreads on CeS and dissolves ∼ 100 atom ppm S, which precipitates out as US on cooldown. Iron wets CeS at 1873 K and 1973 K but does not spread or interact. Because of the desirable containment characteristics of CeS and similar sulfides for molten metals, we recommend their use in a number of applications. (author)

  3. Oxidation of Ce(III) in Foam Decontaminant by Ozone

    International Nuclear Information System (INIS)

    Jung, Chong Hun; Yoon, I. H.; Choi, W. K.; Moon, J. K.; Yang, H. B.; Lee, J. S.

    2016-01-01

    A nanoparticle-based foam decontaminant is composed of a surfactant and nanoparticles for the generation and maintenance of foam, and a chemical decontamination agent made of Ce(IV) dissolved in nitric acid. Ce(IV) will be reduced to Ce(III) through the decontamination process. Oxidizing cerium(III) can be reused as a decontamination agent, Ce(IV). Oxidation treatment technology by ozone uses its strong oxidizing power. It can be regarded as an environmentally friendly process, because ozone cannot be stored and transported like other industrial gases (because it quickly decays into diatomic oxygen) and must therefore be produced on site, and used ozone can decompose immediately. The ozonation treatment of Ce(III) in foam decontaminant containing a surfactant is necessary for the effective regeneration of Ce(III). Thus, the present study was undertaken to determine the optimal conditions for ozonation treatment in the regeneration of Ce(III) into Ce(IV) in the nanoparticle-based foam decontaminant containing surfactant. This study was undertaken to determine the optimal conditions for ozonation treatment in the regeneration of Ce(III) to Ce(IV) in nanoparticle-based foam decontaminant containing a TBS surfactant. The oxidation conversion rate of Ce(III) was increased with an increase in the flow rate of the gas mixture and ozone injection amount. The oxidation time required for the 100% oxidation conversion of Ce(III) to Ce(IV) at a specific ozone injection amount can be predicted from these experimental data

  4. Oxidation of Ce(III) in Foam Decontaminant by Ozone

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chong Hun; Yoon, I. H.; Choi, W. K.; Moon, J. K.; Yang, H. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, J. S. [Gachon University, Seongnam (Korea, Republic of)

    2016-10-15

    A nanoparticle-based foam decontaminant is composed of a surfactant and nanoparticles for the generation and maintenance of foam, and a chemical decontamination agent made of Ce(IV) dissolved in nitric acid. Ce(IV) will be reduced to Ce(III) through the decontamination process. Oxidizing cerium(III) can be reused as a decontamination agent, Ce(IV). Oxidation treatment technology by ozone uses its strong oxidizing power. It can be regarded as an environmentally friendly process, because ozone cannot be stored and transported like other industrial gases (because it quickly decays into diatomic oxygen) and must therefore be produced on site, and used ozone can decompose immediately. The ozonation treatment of Ce(III) in foam decontaminant containing a surfactant is necessary for the effective regeneration of Ce(III). Thus, the present study was undertaken to determine the optimal conditions for ozonation treatment in the regeneration of Ce(III) into Ce(IV) in the nanoparticle-based foam decontaminant containing surfactant. This study was undertaken to determine the optimal conditions for ozonation treatment in the regeneration of Ce(III) to Ce(IV) in nanoparticle-based foam decontaminant containing a TBS surfactant. The oxidation conversion rate of Ce(III) was increased with an increase in the flow rate of the gas mixture and ozone injection amount. The oxidation time required for the 100% oxidation conversion of Ce(III) to Ce(IV) at a specific ozone injection amount can be predicted from these experimental data.

  5. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  6. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    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

  7. The timing of reactor dismantling

    International Nuclear Information System (INIS)

    Roberts, P.

    2000-01-01

    Work has been progressing across the world for the decommissioning of nuclear reactors. The initial work focused on the early, complete dismantling but this was associated with small size reactors and was done for experimental or demonstration purposes. The situation now is that an increasing number of full size power reactors are being shutdown and decision are being made as to the decommissioning strategy to be applied, e.g. with respect to the appropriate timing of reactor dismantling. There are two basic approaches to the timing of reactor dismantling, which are to either proceed with dismantling on an early time scale or to delay it for a period of years. There are a number of examples worldwide of both approaches being taken but one common feature of the approach taken by most countries is that decisions are made on a case by case basis, taking account of relevant factors, and as a result the strategy can vary from reactor to reactor and from country to country. Decisions on timing take account of the following main factors: safety, radioactive decay, financial factors, radioactive waste, reactor type, technology, repository availability, site re-use, regulatory standards, plant knowledge/records, other issues

  8. Calculations of coupled channels for the reaction 142Ce(α,α1)142 Ce*

    International Nuclear Information System (INIS)

    Appoloni, C.R.; Lepine, A.

    1980-01-01

    Elastic and inelastic angular distribution were made for α particles of 18 Mev in 142 Ce. It was determined the angular distributions corresponding to the various states of the target nucleus. The angular distributions corresponding to the first five states were analyzed within the framework of the Anarhmonic vibrational and symmetric rotational models. (A.C.A.S.) [pt

  9. The IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors

  10. Band structures in near spherical 138Ce

    Science.gov (United States)

    Bhattacharjee, T.; Chanda, S.; Bhattacharyya, S.; Basu, S. K.; Bhowmik, R. K.; Das, J. J.; Pramanik, U. Datta; Ghugre, S. S.; Madhavan, N.; Mukherjee, A.; Mukherjee, G.; Muralithar, S.; Singh, R. P.

    2009-06-01

    The high spin states of N=80138Ce have been populated in the fusion evaporation reaction 130Te( 12C, 4n) 138Ce at E=65 MeV. The γ transitions belonging to various band structures were detected and characterized using an array of five Clover Germanium detectors. The level scheme has been established up to a maximum spin and excitation energy of 23 ℏ and 9511.3 keV, respectively, by including 53 new transitions. The negative parity ΔI=1 band, developed on the 6536.3 keV 15 level, has been conjectured to be a magnetic rotation band following a semiclassical analysis and comparing the systematics of similar bands in the neighboring nuclei. The said band is proposed to have a four quasiparticle configuration of [πgh]⊗[. Other band structures are interpreted in terms of multi-quasiparticle configurations, based on Total Routhian Surface (TRS) calculations. For the low and medium spin states, a shell model calculation using a realistic two body interaction has been performed using the code OXBASH.

  11. Peculiarities of the intermediate valence state of Ce in CeM2Si2 (M = Fe, Co, Ni) compounds

    International Nuclear Information System (INIS)

    Koterlyn, M.; Shcherba, I.; Yasnitskii, R.; Koterlyn, G.

    2007-01-01

    The results of thermoelectric power and the electrical resistivity measurements connected with the intermediate valence (IV) of Ce are presented for the compounds CeM 2 Si 2 (M = Fe, Co, Ni) in the temperature range of 4-800 K. It is shown that CeM 2 Si 2 are Kondo-lattices with the coherence scale T coh ∼ 60-80 K and the so-called single-site Kondo temperature T K ∼ 10 3 K. On the example of CeNi 2 Si 2 we have studied the changes in the structure of density of f states (f-DOS) near the Fermi energy caused by atomic substitutions. The results of structural, transport, magnetic, and Ce L III X-ray absorption spectra measurements in the series Ce 1-x La x Ni 2 Si 2 (0 ≤ x ≤ 0.6), Ce(Ni 1-y Cu y ) 2 Si 2 (0 ≤ y ≤ 0.6) and CeNi 2 (Si 1-z Ge z ) 2 (0 ≤ z ≤ 0.5) are presented. We found that the IV state of Ce in the CeM 2 Si 2 is an evidence of possible opening a wide pseudogap Δ ∼ kT K within the f-DOS structure slightly above the Fermi energy

  12. Effect of Co3O4 and Co3O4/CeO2 infiltration on the catalytic and electro-catalytic activity of LSM15/CGO10 porous cells stacks for oxidation of propene

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2015-01-01

    The objective of this work was to study the effect of Co3O4 and Co3O4/CeO2 infiltration on the propene oxidation catalytic activity of a La0.85Sr0.15MnO3/Ce0.9Gd0.1O1.95 electrochemical porous cell stack (11 layers, 5 single cells in series). The effect of the infiltration of Co3O4 and Co3O4/CeO2...... on the electrochemical properties of the porous cell stack was also investigated by electrochemical impedance spectroscopy (EIS). Co3O4 and Co3O4/CeO2 exhibited high catalytic activity for propene oxidation. The increase of propene oxidation rate with +4 V (0.8 V/cell) polarization reached 10% for the Co3O4 infiltrated...... reactor and 48% of efficiency at 300 °C. The Co3O4/CeO2 co-infiltration decreased the reactor polarization resistance, while Co3O4 infiltration had negligible effect on reactor electrochemical performance. The beneficial effect of CeO2 on the electrode activity was attributed to the increased...

  13. CEF-scheme of a semimetal Ce3Sn7

    International Nuclear Information System (INIS)

    Okuda, Yusuke; Yamamoto, Takeshi; Honda, Daisuke; Shishido, Hiroaki; Galatanu, Andrei; Haga, Yoshinori; Matsuda, Tatsuma D.; Takeuchi, Tetsuya; Kindo, Koichi; Sugiyama, Kiyohiro; Settai, Rikio; O-bar nuki, Yoshichika

    2005-01-01

    We measured the magnetic susceptibility and magnetization of an antiferromagnet Ce 3 Sn 7 with the orthorhombic crystal structure. The experimental data are found to be well explained on the basis of the crystalline electric field (CEF) 4f-scheme under the assumption that two Ce atoms in the 2(a) site possess a magnetic moment of 0.36μ B /Ce and one Ce atom in the 4(i) site possesses no magnetic moment as in a valence fluctuating compound CeSn 3 , which was previously proposed by Bonnet et al. Furthermore, we carried out the de Haas-van Alphen experiment. The detected Fermi surfaces are many in number but are extremely small in volume, indicating that Ce 3 Sn 7 is a semimetal

  14. A pressure study of CePt{sub 3}B

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Daniela; Suellow, Stefan [Institute of Condensed Matter Physics, University of Technology Braunschweig, Braunschweig (Germany); Hartwig, Steffen [Institute of Condensed Matter Physics, University of Technology Braunschweig, Braunschweig (Germany); BENSC, Helmholtz Zentrum Berlin, Berlin (Germany); Hidaka, Hiroyuki; Yamazaki, Seigo; Amitsuka, Hiroshi [Department of Physics, Hokkaido University, Sapporo (Japan); Bauer, Ernst [Institute of Solid State Physics, Vienna University of Technology, Vienna (Austria)

    2013-07-01

    CePt{sub 3}B is isostructural to the non-centro symmetric heavy-fermion superconductor CePt{sub 3}Si. In contrast to the latter system, CePt{sub 3}B exhibits a complex magnetically ordered state at low temperatures, with an antiferromagnetic phase below T{sub N}=7.8 K and a weakly ferromagnetic transition below T{sub C}∼5 K. CePt{sub 3}B can be understand as a low pressure variant of CePt{sub 3}Si. Here we report a study of CePt{sub 3}B by means of high pressure magnetization measurements, this way in particular accessing the pressure evolution of the ferromagnetic transition temperature T{sub C}. From our investigation up to about 40 kbar we observe an almost constant transition temperature T{sub C} with pressure. This behavior we discuss in the context of alloying studies on this material.

  15. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  16. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  17. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  18. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-09-01

    The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

  19. The safety features of an integrated maritime reactor

    International Nuclear Information System (INIS)

    Miyakoshi, Junichi; Yamada, Nobuyuki; Kuwahara, Shin-ichi

    1975-01-01

    The EFDR-80, a typical integrated maritime reactor, which is being developed in West Germany is outlined. The safety features of the integrated maritime reactor are presented with the analysis of reactor accidents and hazards, and are compared with those of the separated maritime reactor. Furthermore, the safety criteria of maritime reactors in Japan and West Germany are compared, and some of the differences are presented from the viewpoint of reactor design and safety analysis. In this report the authors express an earnest desire that the definite and reasonable safety criteria of the integrated maritime reactor should be established and that the safety criteria of the nuclear ship should be standardized internationally. (auth.)

  20. Effect of Annealing Time of YAG:Ce3+ Phosphor on White Light Chromaticity Values

    Science.gov (United States)

    Abd, Husnen R.; Hassan, Z.; Ahmed, Naser M.; Almessiere, Munirah Abdullah; Omar, A. F.; Alsultany, Forat H.; Sabah, Fayroz A.; Osman, Ummu Shuhada

    2018-02-01

    Yttrium and aluminium nitrate phosphors doped with cerium nitrate and mixed with urea (fuel) are prepared by using microwave-induced combustion synthesis according to the formula Y(3-0.06)Al5O12:0.06Ce3+ (YAG:Ce3+) to produce white light emitting diodes by conversion from blue indium gallium nitride-light emitting diode chips. The sintering time with fixed temperature (1050°C) for phosphor powder was optimized and found to be 5 h. The crystallinity, structure, chemical composition, luminescent properties with varying currents densities and chromaticity were characterized by x-ray diffraction, field emission-scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, photoluminescence emission, electroluminescence and standard CIE 1931 chromaticity diagram, respectively. The energy levels of Ce3+ in YAG were discussed based on its absorption and excitation spectra. The results show that the obtained YAG:Ce3+ phosphor sintered for 5 h has good crystallinity with pure phase, low agglomerate with spherical shaped particles and strong yellow emission, offering cool-white LED with tuneable correlated color temperature and a good color rendering index compared to those prepared by sintering for 2 h and as-prepared phosphor powders.

  1. Distribution and Translocation of 141Ce (III) in Horseradish

    Science.gov (United States)

    Guo, Xiaoshan; Zhou, Qing; Lu, Tianhong; Fang, Min; Huang, Xiaohua

    2007-01-01

    Background and Aims Rare earth elements (REEs) are used in agriculture and a large amount of them contaminate the environment and enter foods. The distribution and translocation of 141Ce (III) in horseradish was investigated in order to help understand the biochemical behaviour and toxic mechanism of REEs in plants. Method The distribution and translocation of 141Ce (III) in horseradish were investigated using autoradiography, liquid scintillation counting (LSC) and electron microscopic autoradiography (EMARG) techniques. The contents of 141Ce (III) and nutrient elements were analysed using an inductively coupled plasma-atomic emission spectrometer (ICP-AES). Results The results from autoradiography and LSC indicated that 141Ce (III) could be absorbed by horseradish and transferred from the leaf to the leaf-stalk and then to the root. The content of 141Ce (III) in different parts of horseradish was as follows: root > leaf-stalk > leaf. The uptake rates of 141Ce (III) in horseradish changed with the different organs and time. The content of 141Ce (III) in developing leaves was greater than that in mature leaves. The results from EMARG indicated that 141Ce (III) could penetrate through the cell membrane and enter the mesophyll cells, being present in both extra- and intra-cellular deposits. The contents of macronutrients in horseradish were decreased by 141Ce (III) treatment. Conclusions 141Ce (III) can be absorbed and transferred between organs of horseradish with time, and the distribution was found to be different at different growth stages. 141Ce (III) can enter the mesophyll cells via apoplast and symplast channels or via plasmodesmata. 141Ce (III) can disturb the metabolism of macronutrients in horseradish. PMID:17921527

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  3. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  4. Standardized sampling system for reactor coolants

    International Nuclear Information System (INIS)

    Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

    1982-09-01

    A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

  5. Synthesis and structural characterization of Ce-doped bismuth titanate

    International Nuclear Information System (INIS)

    Pavlovic, Nikolina; Srdic, Vladimir V.

    2009-01-01

    Ce-modified bismuth titanate nanopowders Bi 4-x Ce x Ti 3 O 12 (x ≤ 1) have been synthesized using a coprecipitation method. DTA/TG, FTIR, XRD, SEM/EDS and BET methods were used in order to investigate the effect of Ce-substitution on the structure, morphology and sinterability of the obtained powders. The phase structure investigation revealed that after calcinations at 600 deg. C powder without Ce addition exhibited pure bismuth titanate phase; however, powders with Ce (x = 0.25, 0.5 and 0.75) had bismuth titanate pyrochlore phase as the second phase. The strongest effect of Ce addition on the structure was noted for the powder with the highest amount of Ce (x = 1) having a cubic pyrochlore structure. The presence of pure pyrochlore phase was explained by its stabilization due to the incorporation of cerium ions in titanate structure. Ce-modified bismuth titanate ceramic had a density over 95% of theoretical density and the fracture in transgranular manner most probably due to preferable distribution of Ce in boundary region

  6. Inverted opal luminescent Ce-doped silica glasses

    Directory of Open Access Journals (Sweden)

    R. Scotti

    2006-01-01

    Full Text Available Inverted opal Ce-doped silica glasses (Ce : Si molar ratio 1 ⋅ 10−3 were prepared by a sol-gel method using opals of latex microspheres as templates. The rare earth is homogeneously dispersed in silica host matrix, as evidenced by the absence of segregated CeO2, instead present in monolithic Ce-doped SG with the same cerium content. This suggests that the nanometric dimensions of bridges and junctions of the host matrix in the inverted opal structures favor the RE distribution avoiding the possible segregation of CeO2.

  7. CeO2-ZrO2 ceramic compounds

    International Nuclear Information System (INIS)

    Melo, F.C.L.; Cairo, C.A.C.; Devezas, T.C.; Nono, M.C.A.

    1988-01-01

    In order to study the mechanical properties of tetragonal polycrystal zirconia stabilized with ceria various powder compositions with different CeO 2 content were made. Modulus of rupture for those compounds was measured. Tetragonal retained phase was determined for samples of CeO 2 -ZrO 2 ceramics with and without superficial mechanical treatment. The experimental results allowed us to evaluate the effects of CeO 2 content and sintering temperature in the mechanical properties and tetragonal transformed phase (t→ m) in ceramics of CeO 2 -ZrO 2 systems. (author) [pt

  8. The highest spin discrete levels in 131,132Ce

    International Nuclear Information System (INIS)

    Paul, E S; Choy, P T W; Andreoiu, C; Boston, A J; Evans, A O; Fox, C; Gros, S; Nolan, P J; Rainovski, G; Sampson, J A; Scraggs, H C; Walker, A; Appelbe, D E; Joss, D T; Simpson, J; Gizon, J; Astier, A; Buforn, N; Prevost, A; Redon, N; Stezowski, O; Nyako, B M; Sohler, D; Timar, J; Zolnai, L; Bazzacco, D; Lunardi, S; Petrache, C M; Bednarczyk, P; Curien, D; Kintz, N; Ragnarsson, I

    2006-01-01

    The three superdeformed (SD) bands in 132 Ce and the two SD bands in 131 Ce have been extended to higher spin following experiments with the EUROBALL IV spectrometer. The two SD bands in 131 Ce have been linked together. However, despite the relatively high population intensity of the bands (up to 5% of the respective channel), it has not been possible to unambiguously link any of the five SD bands into the low-spin, normally deformed structures of 131,132 Ce

  9. Physics and kinetics of TRIGA reactor

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    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)

  10. Fuel recycling and 4. generation reactors

    International Nuclear Information System (INIS)

    Devezeaux de Lavergne, J.G.; Gauche, F.; Mathonniere, G.

    2012-01-01

    The 4. generation reactors meet the demand for sustainability of nuclear power through the saving of the natural resources, the minimization of the volume of wastes, a high safety standard and a high reliability. In the framework of the GIF (Generation 4. International Forum) France has decided to study the sodium-cooled fast reactor. Fast reactors have the capacity to recycle plutonium efficiently and to burn actinides. The long history of reprocessing-recycling of spent fuels in France is an asset. A prototype reactor named ASTRID could be entered into operation in 2020. This article presents the research program on the sodium-cooled fast reactor, gives the status of the ASTRID project and present the scenario of the progressive implementation of 4. generation reactors in the French reactor fleet. (A.C.)

  11. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Bolton, John A.

    1970-01-01

    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

  12. Backfitting of the FRG reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

    1990-05-01

    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 U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

  13. Comparison of Spectral and Scintillation Properties of LuAP:Ce and LuAP:Ce,Sc Single Crystals

    Science.gov (United States)

    Petrosyan, Ashot G.; Derdzyan, Marina; Ovanesyan, Karine; Shirinyan, Grigori; Lecoq, Paul; Auffray, Etiennette; Kronberger, Matthias; Frisch, Benjamin; Pedrini, Christian; Dujardin, Christophe

    2009-10-01

    Scintillation properties of LuAP:Ce and LuAP:Ce,Sc crystal series were studied under excitation by gamma-rays from a 137Cs source. Both series demonstrated comparable optical quality in terms of underlying absorption at 260 nm, slope of the optical edge and transmission in the range of emission. The light yield of LuAP:Ce crystals measured in 0.2 cm times 0.2 cm times 0.8 cm pixels increases linearly with the Ce concentration reaching at 0.58 at. % 6448 plusmn 322 ph/MeV and 9911 plusmn 496 ph/MeV in the long and in the short directions respectively (the light yield ratio is 65%) and shows no sign of light saturation. The energy resolution is found to depend, among other factors, on the uniformity of Ce concentration within the pixels and is improved to 7.1 plusmn 0.4% (I = 0.2 cm), 9.5 plusmn 0.5% (I = 0.8 cm). Intentional co-doping with Sc + ions was tested and resulted in increase of the Ce distribution coefficient to about 0.3. This enabled to increase the concentration of Ce in LuAP:Ce,Sc crystals up to 0.7 at. %, while conserving high optical quality. In contrast to LuAP:Ce, the light yield in LuAP:Ce,Sc crystals does not increase with Ce concentration, the photo peak being gradually suppressed. The involved mechanisms are discussed basing on measurements of the unit cell volumes, Ce concentration uniformity, x-ray rocking spectra, absorption spectra of pure and variously doped LuAP crystals, and emission spectra under different excitations.

  14. Development of control rod position indicator using seismic-resistance reed switches for integral reactor

    International Nuclear Information System (INIS)

    Yu, Je Yong; Kim, Ji Ho; Huh, Hyung; Choi, Myoung Hwan; Sohn, Dong Seong

    2008-01-01

    The Reed Switch Position Transmitter (RSPT) is used as a position indicator for the control rod in commercial nuclear power plants made by ABB-CE. But this position indicator has some problems when directly adopting it to the integral reactor. The Control Element Drive Mechanism (CEDM) for the integral reactor is designed to raise and lower the control rod in steps of 2mm in order to satisfy the design features of the integral reactor which are the soluble boron free operation and the use of a nuclear heating for the reactor start-up. Therefore the resolution of the position indicator for the integral reactor should be achieved to sense the position of the control rod more precisely than that of the RSPT of the ABB-CE. This paper adopts seismic resistance reed switches to the position indicator in order to reduce the damages or impacts during the handling of the position indicator and earthquake

  15. Russian-American venture designs new reactor

    International Nuclear Information System (INIS)

    Newman, P.

    1994-01-01

    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

  16. Monitor for reactor neutron detector

    International Nuclear Information System (INIS)

    Shirakami, Hisayuki; Shibata, Masatoshi

    1992-01-01

    The device of the present invention judges as to whether a neutron detector is normal or not while considering the change of indication value depending on the power change of a reactor core. That is, the device of the present invention comprises a standard value setting device for setting the standard value for calibrating the neutron detector and an abnormality judging device for comparing the standard value with a measured value of the neutron detector and judging the abnormality when the difference is greater than a predetermined value. The measured value upon initialization of each of the neutron detectors is determined as a quasi-standard value. An average value of the difference between the measured value and the quasi-standard value of a plurality of effective neutron detectors at a same level for the height of the reactor core is multiplied to a power rate based on the reactor core power at a position where the neutron detector is disposed upon calibration. The value obtained by adding the multiplied value and the quasi-standard value is determined as a standard value. The abnormality judging device compares the standard value with the measured value of the neutron detector and, if the difference is greater than a predetermined value, the neutron detector is determined as abnormal. As a result, judgement can be conducted more accurately than conventional cases. (I.S.)

  17. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  18. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  19. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  2. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  3. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  4. C-E productivity improvement program

    International Nuclear Information System (INIS)

    Chiu, C.; Ching, B.; Van Haltern, M.L.

    1984-01-01

    This paper describes the C-E Productivity Improvement Program (CEPIP), which is a computer algorithm for heat rate monitoring and diagnostics for a nuclear power plant. CEPIP uses the pattern recognition technique to identify cause(s) of heat rate degradation. The benefits of such an automated program to the plant performance engineer include early identification of the degrading component(s), provision of necessary economic information (cost of energy loss) to the performance engineer so that he can decide when to repair the degrading components, and identification of operator actions resulting in heat rate degradation (such as closing the valve on the live steam extraction line for the second stage of reheat). In summary, CEPIP improves the overall plant performance by increasing the capacity factor. CEPIP was developed to meet the growing needs of the utilities for an automated performance program. The diagnostic capability has been verified by plant data

  5. Radiative β-decay in 141Ce

    International Nuclear Information System (INIS)

    Rao, K.S.G.; Sanjeeviah, H.

    1982-01-01

    The spectral distribution of the continuous gamma radiation accompanying non-unique first forbidden β-decay of 32 d 141 Ce has been measured in the energy range 200-560 keV with a 4.5 cm x 5.08 cm NaI(Tl) scintillation spectrometer. The source electrons were eliminated using an electromagnet. The raw spectrum was corrected for pile-up, finite energy resolution, Compton electron distribution and geometrical γ-detection efficiency using the method of Liden and Starfelt. The corrected distribution is compared with the direct and detour theories of Lewis and Ford, and Ford and Martin, respectively. Total disagreement between experiment and theory was observed over the entire region of the investigated spectrum. In the energy region from 200 to 350 keV, however, the measured spectrum lies between the direct and detour theories. (orig.)

  6. Dependence of the Ce(iii)/Ce(iv) ratio on intracellular localization in ceria nanoparticles internalized by human cells

    KAUST Repository

    Ferraro, Daniela; Tredici, Ilenia G.; Ghigna, Paolo; Castillio-Michel, Hiram; Falqui, Andrea; Di Benedetto, Cristiano; Alberti, Giancarla; Ricci, Vittorio; Anselmi-Tamburini, Umberto; Sommi, Patrizia

    2017-01-01

    CeO2 nanoparticles (CNPs) have been investigated as promising antioxidant agents with significant activity in the therapy of diseases involving free radicals or oxidative stress. However, the exact mechanism responsible for CNP activity has not been completely elucidated. In particular, in situ evidence of modification of the oxidative state of CNPs in human cells and their evolution during cell internalization and subsequent intracellular distribution has never been presented. In this study we investigated modification of the Ce(iii)/Ce(iv) ratio following internalization in human cells by X-ray absorption near edge spectroscopy (XANES). From this analysis on cell pellets, we observed that CNPs incubated for 24 h showed a significant increase in Ce(iii). By coupling on individual cells synchrotron micro-X-ray fluorescence (μXRF) with micro-XANES (μXANES) we demonstrated that the Ce(iii)/Ce(iv) ratio is also dependent on CNP intracellular localization. The regions with the highest CNP concentrations, suggested to be endolysosomes by transmission electron microscopy, were characterized by Ce atoms in the Ce(iv) oxidation state, while a higher Ce(iii) content was observed in regions surrounding these areas. These observations suggest that the interaction of CNPs with cells involves a complex mechanism in which different cellular areas play different roles.

  7. Dependence of the Ce(iii)/Ce(iv) ratio on intracellular localization in ceria nanoparticles internalized by human cells

    KAUST Repository

    Ferraro, Daniela

    2017-01-09

    CeO2 nanoparticles (CNPs) have been investigated as promising antioxidant agents with significant activity in the therapy of diseases involving free radicals or oxidative stress. However, the exact mechanism responsible for CNP activity has not been completely elucidated. In particular, in situ evidence of modification of the oxidative state of CNPs in human cells and their evolution during cell internalization and subsequent intracellular distribution has never been presented. In this study we investigated modification of the Ce(iii)/Ce(iv) ratio following internalization in human cells by X-ray absorption near edge spectroscopy (XANES). From this analysis on cell pellets, we observed that CNPs incubated for 24 h showed a significant increase in Ce(iii). By coupling on individual cells synchrotron micro-X-ray fluorescence (μXRF) with micro-XANES (μXANES) we demonstrated that the Ce(iii)/Ce(iv) ratio is also dependent on CNP intracellular localization. The regions with the highest CNP concentrations, suggested to be endolysosomes by transmission electron microscopy, were characterized by Ce atoms in the Ce(iv) oxidation state, while a higher Ce(iii) content was observed in regions surrounding these areas. These observations suggest that the interaction of CNPs with cells involves a complex mechanism in which different cellular areas play different roles.

  8. Dopant concentration dependence of radiation-induced positive hysteresis of Ce:GSO and Ce:GSOZ

    International Nuclear Information System (INIS)

    Yanagida, Takayuki; Fujimoto, Yutaka; Watanabe, Kenichi

    2014-01-01

    Positive hysteresis and radiation tolerance to high-dose radiation exposure were investigated for Ce 0.5, 1, and 1.5%-doped Gd 2 SiO 5 (GSO) and for Zr co-doped GSO with the same Ce concentrations (GSOZ). When they were irradiated by 200–800 Gy 60 Co in 200 Gy steps, all Ce-doped GSO samples exhibited light yield enhancement (positive hysteresis). On the other hand, the light yield of GSOZ decreased greatly. Ce 0.5%-doped GSO showed the highest positive hysteresis, with ∼20% light yield enhancement. When the Ce concentration was increased, the positive hysteresis became weaker. - Highlights: • Positive hysteresis Ce 0.5, 1, and 1.5% doped GSO and GSOZ are studied. • Ce 0.5, 1, and 1.5% doped GSO show the positive hysteresis by 2–8 M rad 60 Co irradiation. • Ce 0.5, 1, and 1.5% doped GSOZ do not show the positive hysteresis. • By Zn co-doping, radiation tolerance of GSO becomes weaker. • By dense Ce doping, radiation tolerance of GSO and GSOZ are improved

  9. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  10. Applications of Research Reactors

    International Nuclear Information System (INIS)

    2014-01-01

    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

  11. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  12. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  13. Preparation of catalysts based on Ce-Mn mixed oxide by coprecipitation for combustion of n-hexane

    International Nuclear Information System (INIS)

    Picasso, Gino; Zavala, Cesar; Cruz, Romulo; Sun Kou, Rosario; Lopez, Alcides

    2013-01-01

    Catalysts based on Ce-Mn mixed with different Ce/Mn molar ratios ranging from 0,5 to 2 have been prepared by coprecipitation at pH constant with ageing times of 4, 18 and 24 h for combustion of n-hexane. XRD patterns of the mixed oxides showed the majority presence of fluorite phase. Specific BET surface areas of mixed oxides were always higher than their single counterparts and their adsorption isotherm depicted a mesoporous surface of Type IV. TPR thermograms confirmed the presence of mixed oxide phase, whose profile shifted to smaller temperatures with increasing content of ceria. Catalytic tests were performed with 2000 ppm of n-hexane and WHSV of 80 h -1 in a fixed-bed reactor. For all samples, only CO 2 and water were observed at total conversion and no partial combustion products were obtained. Ce-Mn mixed oxides were more active than simple oxide samples no matter the aging time. Mixed samples presented thermal stability in contrast with simple ones. Mixed sample with Ce/Mn molar ratio of 2 depicted the highest activity probably due to higher surface area and better reducibility ability of mixed phase. (author)

  14. Traducerea: între ce se poate traduce și ce trebuie tradus

    Directory of Open Access Journals (Sweden)

    Magda Jeanrenaud

    2016-02-01

    Full Text Available Pornind de la o tulburătoare interpretare a lui Jacques Derrida, studiul de față își propune să investigheze și încearcă să explice blocajul ce intervine în versiunile englezești, franceze și românești (semnate de Antoine Berman, Alexis Nouss, Steven Rendall, Catrinel Pleșu etc. ale celebrului text al lui Walter Benjamin, Die Aufgabe des Übersetzers, atunci cînd traducătorii transpun în cele trei limbi țintă cele două citate cuprinse în acesta: un citat din Mallarmé, lăsat netradus de Benjamin însuși, și un altul, din Pannwitz. Într-un fel sau altul, ambele citate au o formă discursivă ce lasă să se întrevadă o sintaxă ce se abate deliberat de la normă, ca și cum ar fi deja niște „traduceri”. Analiza mai pune în evidență și comportamentul (cumva o dominantă a psihologiei traducătorilor? celor ce au transpus textul benjaminian, comportament marcat de obsesia lizibilității văzută ca o trăsătură congenitală a oricărei traduceri, chiar și atunci cînd textul original nu tinde spre aceasta. De unde și dilema, dureroasă, legată de spinoasa chestiune a intenționalității textului (nu doar de tradus...

  15. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  16. Evidence of complex magnetism in CePt3C

    International Nuclear Information System (INIS)

    Vejpravova, J.; Prokleska, J.; Danis, S.; Uhlirova, K.; Sechovsky, V.

    2006-01-01

    CePt 3 C has been synthesized and studied by powder X-ray diffraction (at RT), specific heat, resistivity and magnetization measurements at temperatures from RT down to 0.4K and in magnetic fields up to 10T. A possible scenario for the ground state of CePt 3 C based on the observed phenomena is proposed

  17. 32 CFR Appendix B to Part 247 - CE Publications

    Science.gov (United States)

    2010-07-01

    ... dining at a restaurant or attending a musical performance) of a commercial organization whose primary... potentially become the CE contractor. Upon evaluation of the competing proposals by the Source Selection Advisory Committee (SSAC) and selection of a winner by the selecting official, the CE contract shall be...

  18. Ce que nous faisons | CRDI - Centre de recherches pour le ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Le CRDI appuie des travaux de recherche dans les pays en voie de développement en vue de produire un changement réel et durable. Ce savoir peut servir d'outil pour résoudre des problèmes mondiaux urgents. Nous partageons ce savoir avec les autres en :

  19. Facile hydrothermal synthesis of CeO2 nanopebbles

    Indian Academy of Sciences (India)

    Administrator

    However, to the best of our knowledge the reports on the synthesis of CeO2 ... The base pressure of the XAS chamber was in the range of 10–8 Pa. A Shimadzu ... scopy was investigated to confirm the crystalline quality of CeO2 nanopebbles.

  20. Ferroelectric relaxor Ba(TiCe)O3

    International Nuclear Information System (INIS)

    Chen Ang; Zhi Jing; Yu Zhi

    2002-01-01

    The dielectric behaviour of Ba(Ti 1-y Ce y )O 3 solid solutions (y=0-0.3) has been studied. A small amount of Ce doping (y=0.02) has weak influence on the dielectric behaviour of Ba(Ti 1-y Ce y )O 3 . With increasing Ce concentration, three phase transitions of pure BaTiO 3 are pinched into one rounded dielectric peak with frequency dispersion, and the relaxation time follows the Vogel-Fulcher relation. The evolution from a normal ferroelectric to a ferroelectric relaxor is emphasized. High strains (S=∼0.1-0.19%) with a small hysteresis under ac fields are obtained in ferroelectric relaxors Ba(Ti 1-y Ce y )O 3 . The physical mechanism of the relaxation process, the pinching effect of the phase transitions and their influence on the ferroelectric and electrostrictive behaviour are discussed. (author)

  1. Photodynamic Processes in Fluoride Crystals Doped with Ce3+

    Directory of Open Access Journals (Sweden)

    Pavlov V.V.

    2015-01-01

    Full Text Available Integrated studies of photoelectric phenomena and their associated photodynamic processes in LiCaAlF6, LiLuF4, LiYF4, LiY0,5Lu0,5F4, SrAlF5 crystals doped with Ce3+ ions have been carried out using the combination of the methods of optical and dielectric spectroscopy. The numerical values of the basic parameters of photodynamic processes and their spectral dependence in 240 – 310 nm spectral range are evaluated. It has been shown that the most probable process, which leads to the photoionization of Ce3+ ions in LiYxLu1-xF4:Ce3+ (x=0; 0,5; 1 and LiCaAlF6:Ce3+ crystals, is excited-state absorption to the states of mixed configurations of Ce3+ ions localized near/in the conduction band of crystal.

  2. Instrumentation qualification. Seismic qualification of C-E instrumentation equipment. Part One

    International Nuclear Information System (INIS)

    1977-05-01

    A summary of the C-E seismic qualification program utilized to demonstrate the seismic design adequacy of the instrumentation and control equipment used in C-E supplied Nuclear Steam Supply Systems (NSSS) is presented. The report is divided into two parts. Part One includes the equipment seismic requirements and a description of the qualification methods. Part Two lists the specific equipment by nuclear station in which it is used and the equipment test results are summarized in a standard data sheet format to facilitate review. The seismic requirements are based on individual contract commitments with C-E customers and the NRC Standard Review Plan, Section 3.10 ''Seismic Qualification of Category I Instrumentation and Electrical Equipment.'' Equipment is qualified for use in a seismic environment where damage potential to the equipment is less than or equal to that simulated seismic environment to which it has been qualified. The anticipated Safe Shutdown Earthquake (SSE) environment at the inservice location of equipment should be confirmed by each applicant as not exceeding that to which it is qualified

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  7. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  8. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  9. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  12. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  13. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  14. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  16. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  17. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  18. A Novel Assembly Line Scheduling Algorithm Based on CE-PSO

    Directory of Open Access Journals (Sweden)

    Xiaomei Hu

    2015-01-01

    Full Text Available With the widespread application of assembly line in enterprises, assembly line scheduling is an important problem in the production since it directly affects the productivity of the whole manufacturing system. The mathematical model of assembly line scheduling problem is put forward and key data are confirmed. A double objective optimization model based on equipment utilization and delivery time loss is built, and optimization solution strategy is described. Based on the idea of solution strategy, assembly line scheduling algorithm based on CE-PSO is proposed to overcome the shortcomings of the standard PSO. Through the simulation experiments of two examples, the validity of the assembly line scheduling algorithm based on CE-PSO is proved.

  19. [18F]DOPA PET/ceCT in diagnosis and staging of primary medullary thyroid carcinoma prior to surgery.

    Science.gov (United States)

    Rasul, Sazan; Hartenbach, Sabrina; Rebhan, Katharina; Göllner, Adelina; Karanikas, Georgios; Mayerhoefer, Marius; Mazal, Peter; Hacker, Marcus; Hartenbach, Markus

    2018-05-15

    Medullary thyroid carcinoma (MTC) is characterized by a high rate of metastasis. In this study we evaluated the ability of [ 18 F]DOPA PET/ceCT to stage MTC in patients with suspicious thyroid nodules and pathologically elevated serum calcitonin (Ctn) levels prior to total thyroidectomy and lymph node (LN) dissection. A group of 32 patients with sonographically suspicious thyroid nodules and pathologically elevated basal Ctn (bCtn) and stimulated Ctn (sCtn) levels underwent DOPA PET/ceCT prior to surgery. Postoperative histology served as the standard of reference for ultrasonography and DOPA PET/ceCT region-based LN staging. Univariate and multivariate regression analyses as well as receiver operating characteristic analysis were used to evaluate the correlations between preoperative and histological parameters and postoperative tumour persistence or relapse. Primary MTC was histologically verified in all patients. Of the 32 patients, 28 showed increased DOPA decarboxylase activity in the primary tumour (sensitivity 88%, mean SUVmax 10.5). Undetected tumours were exclusively staged pT1a. The sensitivities of DOPA PET in the detection of central and lateral metastatic neck LN were 53% and 73%, in contrast to 20% and 39%, respectively, for neck ultrasonography. Preoperative bCtn and carcinoembryonic antigen levels as well as cN1b status and the number of involved neck regions on DOPA PET/ceCT were predictive of postoperative tumour persistence/relapse in the univariate regression analysis (P PET/ceCT cN1b status remained significant in the multivariate analysis (P = 0.016, relative risk 4.02). This study revealed that DOPA PET/ceCT has high sensitivity in the detection of primary MTC and superior sensitivity in the detection of LN metastases compared to ultrasonography. DOPA PET/ceCT identification of N1b status predicts postoperative tumour persistence. Thus, implementation of a DOPA-guided LN dissection might improve surgical success.

  20. Archiver ce qui aurait pu avoir lieu

    Directory of Open Access Journals (Sweden)

    Stefanie Baumann

    2009-12-01

    Full Text Available L’Atlas Group, un projet de l’artiste libanais Walid Raad, est « dédié à la recherche et la compilation de documents sur l’histoire contemporaine libanaise. L’Atlas Group produit, localise, conserve et étudie des documents visuels, sonores, textuels et autres, qui mettent en lumière l’histoire actuelle du Liban. »Ce projet est ainsi présenté comme une fondation qui génère des archives historiques et qui collecte des traces relevant de la guerre au Liban afin de les mettre à disposition aux chercheurs. Mais, les matériaux sont produits par l’artiste : l’archive est imaginaire, les documents et récits sont inventés, ainsi que le Docteur Fakhouhi, le personnage principal, présenté comme étant « le plus renommé des historiens au Liban ». Le spectateur se trouve alors devant un scénario très étrange dans lequel sont détournées les notions de document (qui peut désormais être aussi bien trouvé que produit et d’histoire (car les situations décrites sont considérées comme « ayant très bien pu avoir eu lieu », l’Atlas Group  les traite comme de véritables événement historiques et qui déplace, mine de rien, tout un dispositif. Au sein de ce projet, la frontière entre fiction et documentaire est complètement estompée : le geste de l’artiste (qui, lui, se présente comme une institution, un « Groupe », interroge ainsi le statut même d’auteur vise à la déplacer pour poser des questions relatives aux représentations possibles de l’histoire, aux personnes aptes à se charger de son écriture et à l’opération historiographique.

  1. Temperature etalon of WWER-440 reactor

    International Nuclear Information System (INIS)

    Stanc, S.; Slanina, M.

    2001-01-01

    The presentation deals with the description, parameters and advantages of use of the temperature etalon. The system ensures temperature measurement of reactor outlet and inlet temperatures with high accuracy. Accuracy of temperature measurement is 0.18 deg C, accuracy of temperature difference measurement is 0.14 deg C, both with probability 0.95. Using the temperature etalon it is possible to increase accuracy of the standard temperature reactor measurements and to check their accuracy in the course of power reactor statuses in every measurement cycle. Temperature reactor etalon was installed in 12 WWER-440 units in Slovakia, Bohemia and Bulgaria. (Authors)

  2. Tetragonal zirconia ceramics in Zr O2-Ce O2 system (Ce-TZP): preparation, characterization and mechanical properties

    International Nuclear Information System (INIS)

    Andrade Nono, M.C. de.

    1992-01-01

    This paper describes and discusses the results achieved in a study about Ce-TZP ceramics prepared from conventional powder mixtures of Zr O 2 and Ce O 2 (with composition in the range of 8 to 16 mol% Ce O 2 ). Physical and chemical characteristics were related with the powder compaction behavior and with the sintering state. The sintered ceramics showed a level of high porosity (≅ 4%), mainly due to the fairly adequate powder characteristics and compaction. The crystalline phases were analysed from X-rays diffraction data and showed that these ceramics can present tetragonal-to-monoclinic stress induced transformation. The bending strength, fracture toughness and Vickers hardness results were influenced by Ce O 2 content microstructure and sintering temperature. These Ce-TZP ceramics showed mechanical strength results comparable to those published in the international literature. (author)

  3. International Electrotechnical Commission standards and French material control standards

    International Nuclear Information System (INIS)

    Furet, J.; Weill, J.

    1978-01-01

    There are reported the international standards incorporated into the IEC Subcommitee 45 A (Nuclear Reactor Instrumentation) and the national standards elaborated by the Commissariat a l'Energie Atomique, CEA, Group of normalized control equipment, the degree of application of those being reported on the base design, call of bids and exploitation of nuclear power plants. (J.E. de C)

  4. The System 80+ standard plant design reduces operations and maintenance costs

    International Nuclear Information System (INIS)

    Chari, D.R.; Robertson, J.E.

    1998-01-01

    To be cost-competitive, nuclear power plants must maximize plant availability and minimize operations and maintenance (O and M) costs. A plant whose design supports these goals will generate more power at less cost and thereby have a lower unit generating cost. The ABB Combustion Engineering Nuclear Systems (ABB-CE) System 80+ Standard Nuclear Power Plant, rated at 1400 megawatts electric (MWe), is designed for high availability at reduced cost. To demonstrate that the duration of refueling outages, the major contributor to plant unavailability, can be shortened, ABB-CE developed a detailed plan that shows a System 80+ plant can safely perform a refueling and maintenance outage in 18 days. This is a significant reduction from the average current U.S. plant outages of 45 days, and is possible due to a two-part outage strategy: use System 80+ advanced system design features and relaxed technical specification (TS) time limits to shift some maintenance from outages to operating periods: and, use System 80+ structural, system, and component features, such as the larger operating floor, permanent pool seal, integral reactor head area cable tray system and missile shield, and longer life reactor coolant pump seals, to reduce the scope and duration of outage maintenance activities. Plant staffing level is the major variable, or controllable contributor to operations costs. ABB-CE worked with the Institute of Nuclear Power Operations (INPO) to perform detailed staffing analyses that show a System 80+ plant can be operated reliably with 30 percent less staff than currently operating nuclear plants of similar size. Safety was not sacrificed when ABB-CE developed the System 80+ refueling outage plan and staffing level. The outage plan was developed utilizing a defense-in-depth concept for shutdown safety. The defense in-depth concept is implemented via systematic control of outage risk evaluation (SCORE) cards. The SCORE cards identify primary and alternate means of

  5. Fermi surface study of CeSb

    International Nuclear Information System (INIS)

    Aoki, H.; Crabtree, G.W.; Joss, W.; Hulliger, F.

    1984-09-01

    A Fermi surface study of the ferromagnetic phase of CeSb is presented. The γ frequency branches arising from the electron surfaces at the X points, three separate frequency branches from the hole surfaces at the GAMMA point and the low frequency branch α have been observed. The effective mass ratios are low and range from approx. 0.2 for the α branch to approx. 1.0 for the high frequency branch of γ. The low effective mass ratios suggest that the admixture of the conduction states with the f state is small. We have observed a drastic change in the appearance of the dHvA signal at the phase transition between the ferromagnetic and lower field antiferromagnetic phases: The low frequency α oscillation suddenly disappears as the crystal enters the antiferromagnetic phase. By utilizing the change in the signal appearance, the transition field strength has been measured as a function of the field direction. The present experimental results, particularly the origin of the α oscillation, are discussed in the light of the p-f mixing theory and recent band structure calculations based on localized f orbitals

  6. Fermi surface study of CeSb

    International Nuclear Information System (INIS)

    Aoki, H.; Crabtree, G.; Joss, W.; Hulliger, F.

    1985-01-01

    A Fermi surface study of the ferromagnetic phase of CeSb is presented. The γ frequency branches arising from the electron surfaces at the X points, three separate frequency branches from the hole surfaces at the GAMMA point, and the low-frequency branch α have been observed. The effective mass ratios are low and range from approx.0.2 for the α branch to approx.1.0 for the high-frequency branch of γ. The low effective mass ratios suggest that the admixture of the conduction states with the f state is small. We have observed a drastic change in the appearance of the de Haas--van Alpen signal at the phase transition between the ferromagnetic and lower field antiferromagnetic phases: the low-frequency α oscillation suddenly disappears as the crystal enters the antiferromagnetic phase. By utilizing the change in the signal appearance, the transition field strength has been measured as a function of the field direction. The present experimental results particularly the origin of the α oscillation, are discussed in the light of the p-f mixing theory and recent band-structure calculations based on localized f orbitals

  7. Recent approaches in sensitive enantioseparations by CE.

    Science.gov (United States)

    Sánchez-Hernández, Laura; Castro-Puyana, María; Marina, María Luisa; Crego, Antonio L

    2012-01-01

    The latest strategies and instrumental improvements for enhancing the detection sensitivity in chiral analysis by CE are reviewed in this work. Following the previous reviews by García-Ruiz et al. (Electrophoresis 2006, 27, 195-212) and Sánchez-Hernández et al. (Electrophoresis 2008, 29, 237-251; Electrophoresis 2010, 31, 28-43), this review includes those papers that were published during the period from June 2009 to May 2011. These works describe the use of offline and online sample treatment techniques, online sample preconcentration techniques based on electrophoretic principles, and alternative detection systems to UV-Vis to increase the detection sensitivity. The application of the above-mentioned strategies, either alone or combined, to improve the sensitivity in the enantiomeric analysis of a broad range of samples, such as pharmaceutical, biological, food and environmental samples, enables to decrease the limits of detection up to 10⁻¹² M. The use of microchips to achieve sensitive chiral separations is also discussed. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  9. Measurement of home-made LaCl3 : Ce scintillation detector sensitivity with different energy points in range of fission energy

    International Nuclear Information System (INIS)

    Hu Mengchun; Li Rurong; Si Fenni

    2010-01-01

    Gamma rays of different energy were obtained in the range of fission energy by Compton scattering in intense 60 Co gamma source and the standard isotopic gamma sources which are 0.67 MeV 137 Cs and l.25 MeV 60 Co sources of point form. Sensitivity of LaCl 3 : Ce scintillator was measured in these gamma ray energy by a fast response scintillation detector with the home-made LaCl 3 : Ce scintillator. Results were normalized by the sensitivity to 0.67 MeV gamma ray. Sensitivity of LaCl 3 : Ce to 1.25 MeV gamma ray is about l.28. For ø40 mm × 2 mm LaCl 3 : Ce scintillator, the biggest sensitivity is l.18 and the smallest is 0.96 with gamma ray from 0.39 to 0.78 MeV. And for ø40 mm × 10 mm LaCl 3 : Ce scintillator, the biggest sensitivity is l.06 and the smallest is 0.98. The experimental results can provide references for theoretical study of the LaCl 3 : Ce scintillator and data to obtain the compounded sensitivity of LaCl 3 : Ce scintillator in the range of fission energy. (authors)

  10. Safety-related parameters for the MAPLE research reactor and a comparison with the IAEA generic 10-MW research reactor

    International Nuclear Information System (INIS)

    Carlson, P.A.; Lee, A.G.; Smith, H.J.; Ellis, R.J.

    1989-07-01

    A summary is presented of some of the principle safety-related physics parameters for the MAPLE Research Reactor, and a comparison with the IAEA Generic 10-MW Reactor is given. This provides a means to assess the operating conditions and fuelling requirements for safe operation of the MAPLE Research Reactor under accepted standards

  11. Hydrogen separation through tailored dual phase membranes with nominal composition BaCe0.8Eu0.2O3-δ:Ce0.8Y0.2O2-δ at intermediate temperatures

    Science.gov (United States)

    Ivanova, Mariya E.; Escolástico, Sonia; Balaguer, Maria; Palisaitis, Justinas; Sohn, Yoo Jung; Meulenberg, Wilhelm A.; Guillon, Olivier; Mayer, Joachim; Serra, Jose M.

    2016-11-01

    Hydrogen permeation membranes are a key element in improving the energy conversion efficiency and decreasing the greenhouse gas emissions from energy generation. The scientific community faces the challenge of identifying and optimizing stable and effective ceramic materials for H2 separation membranes at elevated temperature (400-800 °C) for industrial separations and intensified catalytic reactors. As such, composite materials with nominal composition BaCe0.8Eu0.2O3-δ:Ce0.8Y0.2O2-δ revealed unprecedented H2 permeation levels of 0.4 to 0.61 mL·min-1·cm-2 at 700 °C measured on 500 μm-thick-specimen. A detailed structural and phase study revealed single phase perovskite and fluorite starting materials synthesized via the conventional ceramic route. Strong tendency of Eu to migrate from the perovskite to the fluorite phase was observed at sintering temperature, leading to significant Eu depletion of the proton conducing BaCe0.8Eu0.2O3-δ phase. Composite microstructure was examined prior and after a variety of functional tests, including electrical conductivity, H2-permeation and stability in CO2 containing atmospheres at elevated temperatures, revealing stable material without morphological and structural changes, with segregation-free interfaces and no further diffusive effects between the constituting phases. In this context, dual phase material based on BaCe0.8Eu0.2O3-δ:Ce0.8Y0.2O2-δ represents a very promising candidate for H2 separating membrane in energy- and environmentally-related applications.

  12. C-E setpoint methodology. C-E local power density and DNB LSSS and LCO setpoint methodology for analog protection systems

    International Nuclear Information System (INIS)

    1976-04-01

    A description is presented of the methodology presently in use by Combustion Engineering to calculate Limiting Safety System Setting (LSSS) for the Local Power Density and Thermal Margin Trip Systems and Limiting Conditions for Operation (LCO) to assure that the specified acceptable fuel design limits are not exceeded during the design basis anticipated operational occurrences. The C-E Nuclear Steam Supply Systems for which the report is applicable are those incorporating the analog reactor protection system and licensed under the requirements of 10CFR50, Appendix A. The design basis events to be accommodated by the subject LSSS and LCO are discussed, and the methods to assure the required protection system response and initial required margin are described. The calculational techniques used to represent the specified acceptable fuel design limits in terms of monitored reactor parameters are provided. Using the resultant limits as a base, the methodology to synthesize the subject LSSS and LCO in terms of the parameters processed by the protection and monitoring systems is described

  13. Mechanochemical and combustion synthesis of CeB{sub 6}

    Energy Technology Data Exchange (ETDEWEB)

    Akguen, Baris [Roketsan Missiles Inc., Ankara (Turkey); Sevinc, Naci; Topkaya, Yavuz [Middle East Technical Univ., Ankara (Turkey). Dept. of Metallurgical and Materials Engineerung; Camurlu, H. Erdem [Akdeniz Univ., Antalya (Turkey). Dept. of Mechanical Engineering

    2013-04-15

    CeB{sub 6} powder was prepared via combustion synthesis (CS) and mechanochemical processing (MCP) methods starting from CeO{sub 2}, B{sub 2}O{sub 3} and Mg powder mixtures. In CS, reactant mixtures were ignited in a preheated pot furnace under argon atmosphere. Products contained CeB{sub 6}, MgO and Mg{sub 3}B{sub 2}O{sub 6}, as revealed by X-ray diffraction analysis. After leaching in 1 M HCl for 15h, MgO was removed but Mg{sub 3}B{sub 2}O{sub 6} could not be removed from the products. Ball milling of products in ethanol prior to leaching made the removal of Mg{sub 3}B{sub 2}O{sub 6} possible by leaching. Yield of CeB{sub 6} was 68.6% in CS. MCP was performed in a stainless steel vial with a planetary ball mill at 300 rpm for 30h. MCP products contained CeB{sub 6}, MgO and small amount of Fe. Leaching in 1 M HCl for 30min was sufficient to remove MgO. Yield of CeB{sub 6} was 84.4% in MCP. According to scanning electron microscopy examinations, particles of CeB6 prepared by CS and MCP had submicrometer size. Average particle sizes were determined as 290nm and 240nm, respectively.

  14. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

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

    International Nuclear Information System (INIS)

    Fonseca, F. de A.S. da; Silva Filho, E.

    1988-12-01

    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) [pt

  16. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-01-01

    This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

  17. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  18. The Simulator Development for RDE Reactor

    Science.gov (United States)

    Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

  19. Safety features of TR-2 reactor

    International Nuclear Information System (INIS)

    Tuerker, T.

    2001-01-01

    TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

  20. A new fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Sefidvash, F.

    1986-01-01

    A new nuclear reactor design based on the fluidized bed concept is proposed. A current design utilizes spherical fuel of slightly enriched Zircaloy-clad uranium dioxide fluidized by light water under pressure. The reactor is modular in system; therefore, any size reactor can be constructed from the basic standard modul. The reactor physics calculations show that reactivity increases with porosity to a maximum value and thereafter decreases. This produces inherent safety and eliminates the need for control rods and burnable poisons. The heat transfer calculations show that the maximum power extracted from the reactor core is not limited to the material temperature limits but to the maximum mass flow of coolant, which corresponds to the desired operating porosity. Design simplicity and inherent safety make it an attractive small reactor design. (Author) [pt

  1. The formation of intermetallic compounds during interdiffusion of Mg–Al/Mg–Ce diffusion couples

    International Nuclear Information System (INIS)

    Dai, Jiahong; Jiang, Bin; Li, Xin; Yang, Qingshan; Dong, Hanwu; Xia, Xiangsheng; Pan, Fusheng

    2015-01-01

    Graphical abstract: Al–Ce intermetallic compounds (IMCs) formed in Mg–Al/Mg–Ce diffusion couples. During the whole diffusion process, Al was the dominant diffusing species, and it substituted for Mg atoms of the Mg–Ce substrate. Five Al–Ce IMCs of Al 4 Ce, Al 11 Ce 3 , Al 3 Ce, Al 2 Ce, and AlCe were formed via the reaction of Al and Ce. - Highlights: • Al–Ce IMCs formation in the Mg–Al/Mg–Ce diffusion couples was studied. • Formation of Al 4 Ce as the first phase was rationalized using the Gibbs free energy. • The activation energy for the growth of the diffusion reaction zones was 36.6 kJ/mol. - Abstract: The formation of Al–Ce intermetallic compounds (IMCs) during interdiffusion of Mg–Al/Mg–Ce diffusion couples prepared by solid–liquid contact method was investigated at 623 K, 648 K and 673 K for 24 h, 48 h and 72 h, respectively. During the whole diffusion process, Al was the dominant diffusing species, and it substituted for Mg of the Mg–Ce substrate. Five Al–Ce IMCs of Al 4 Ce, Al 11 Ce 3 , Al 3 Ce, Al 2 Ce and AlCe were formed via the reaction of Al and Ce. The formation of Al 4 Ce as the first kind of IMC was rationalized on the basis of an effective Gibbs free energy model. The activation energy for the growth of the total diffusion reaction layer was 36.6 kJ/mol

  2. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    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.) [pt

  3. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  4. Properties and practical application of thin CeOx films

    Directory of Open Access Journals (Sweden)

    Maksimchuk N. V.

    2010-10-01

    Full Text Available The properties of CeOx films produced by various methods have been investigated. According to the comparative analisys “metallic mirror oxidation” method allows to produce films with significantly better characteristics than the «explosive evaporation» method. Though the latter method yields higher photosensitivity of CeOx films and structures on their base. In the process the optimal value of the substrate temperature was determined. Obtained data expand the CeOx application potential in microelectronic sensor sphere.

  5. Biological reduction-deposition and luminescent properties of nanostructured CePO{sub 4}@NaCe(SO{sub 4}){sub 2}(H{sub 2}O) and CePO{sub 4}

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Xiaoniu [School of Materials Science and Engineering, Southeast University, Nanjing 211189 (China); Research Institute of Green Construction Materials, Southeast University, Nanjing 211189 (China); Qian, Chunxiang, E-mail: cxqian@seu.edu.cn [School of Materials Science and Engineering, Southeast University, Nanjing 211189 (China); Research Institute of Green Construction Materials, Southeast University, Nanjing 211189 (China)

    2016-03-01

    Nano-sized CePO{sub 4}@NaCe(SO{sub 4}){sub 2}(H{sub 2}O) and CePO{sub 4} with hexagonal phase have been prepared by simply varying the reactant P/Ce molar ratio in bacterial liquid. The phase composition of two samples was checked via Fourier transform infrared spectroscopy (FTIR), energy dispersive analysis of X-rays (EDS) and X-ray diffraction (XRD) analyses, displaying the presence of CePO{sub 4}@NaCe(SO{sub 4}){sub 2}(H{sub 2}O) and CePO{sub 4} with average crystallite size are 32.34 and 15.61 nm, respectively. The scanning electron microscopy (SEM) images show that nano-clusters and sphere-like in shape with a narrow diameter distribution were observed in two samples. The transmission electron microscopy (TEM) photographs further indicate obtained CePO{sub 4}@NaCe(SO{sub 4}){sub 2}(H{sub 2}O) and CePO{sub 4} nanoparticles correspond to nanosheets and nanorods, respectively. The emission spectra of CePO{sub 4}@NaCe(SO{sub 4}){sub 2}(H{sub 2}O) and CePO{sub 4} display a broad band of 300–380 nm range with the strongest emission at 342 nm in the violet region. - Highlights: • A new method was found to synthesize CePO{sub 4}@NaCe(SO{sub 4}){sub 2}(H{sub 2}O) and CePO{sub 4} nanoparticles. • CePO{sub 4}@NaCe(SO{sub 4}){sub 2}(H{sub 2}O) nanoparticles have good luminescent properties. • Size and luminescent properties of two samples have been studied and compared.

  6. HumanViCe: Host ceRNA network in virus infected cells in human

    Directory of Open Access Journals (Sweden)

    Suman eGhosal

    2014-07-01

    Full Text Available Host-virus interaction via host cellular components has been an important field of research in recent times. RNA interference mediated by short interfering RNAs and microRNAs (miRNA, is a widespread anti-viral defence strategy. Importantly, viruses also encode their own miRNAs. In recent times miRNAs were identified as key players in host-virus interaction. Furthermore, viruses were shown to exploit the host miRNA networks to suite their own need. The complex cross-talk between host and viral miRNAs and their cellular and viral targets forms the environment for viral pathogenesis. Apart from protein-coding mRNAs, non-coding RNAs may also be targeted by host or viral miRNAs in virus infected cells, and viruses can exploit the host miRNA mediated gene regulatory network via the competing endogenous RNA effect. A recent report showed that viral U-rich non-coding RNAs called HSUR, expressed in primate virus herpesvirus saimiri (HVS infected T cells, were able to bind to three host miRNAs, causing significant alteration in cellular level for one of the miRNAs. We have predicted protein coding and non protein-coding targets for viral and human miRNAs in virus infected cells. We identified viral miRNA targets within host non-coding RNA loci from AGO interacting regions in three different virus infected cells. Gene ontology (GO and pathway enrichment analysis of the genes comprising the ceRNA networks in the virus infected cells revealed enrichment of key cellular signalling pathways related to cell fate decisions and gene transcription, like Notch and Wnt signalling pathways, as well as pathways related to viral entry, replication and virulence. We identified a vast number of non-coding transcripts playing as potential ceRNAs to the immune response associated genes; e.g. APOBEC family genes, in some virus infected cells. All these information are compiled in HumanViCe, a comprehensive database that provides the potential ceRNA networks in virus

  7. Russian seismic standards and demands for equipment and their conformity with international standards

    International Nuclear Information System (INIS)

    Kaznovsky, S.; Ostretsov, I.

    1993-01-01

    The principle regulations of standard documents concerning seismic safety of NPPs and demands for reactor equipment conformity with international standards are presented in this report. General state of NPP safety standards is reviewed, with a special emphasis on the state of seismic design standards for NPP equipment and piping. Russian standards documents on seismic resistance of NPPs and requirements are compared to international ones

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  10. A review of the Italian fast reactor programme

    Energy Technology Data Exchange (ETDEWEB)

    Bruzzi, L; Pierantoni, F [CNEN Fast Reactor Programme, Bologna (Italy)

    1981-05-01

    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.

  11. Neutron Activation Resonance Integrals of 74Se, 78Se, 80Se, 81Br, 127I, 130Te, 138Ba, 140Ce, and 142Ce

    International Nuclear Information System (INIS)

    Ricabarra, M. D.; Turjanskl, R.; Ricabarra, G. H.; Bigham, C.B.

    1968-01-01

    A lithium-drift germanium γ-ray spectrometer has been used to make accurate intercomparisons of the ratio of resonance-integral to thermal-activation cross section by measuring cadmium ratios or relative activation rates in two different neutron spectra. The standard, gold, or secondary standard, indium, was mixed uniformly in the samples and the activities resolved with the spectrometer. Expressed as Westcott S 0 values, the results relative to S 0 = 17.7 for gold were as follows: 74 Se = 10.3 +± 0.1, 78 Se = 12.3 ± 0.3, 80 Sc = 2.65 ± 0.02, 81 Br = 24.3 ± 0.5, 127 I = 27.8 ± 0.5, 130 Te = 2.10 ± 0.07, 138 Ba = 0.649 ± 0.004, 140 Ce = 0.476 ± 0.003, 142 Ce = 0.865 ± 0.005. (author)

  12. Kinetic study of the ethene oxidation by oxygen in the presence of carbon dioxide and steam over Pt/Rh/CeO2/g-Al2O3

    NARCIS (Netherlands)

    Nibbelke, R.H.; Kreijveld, R.J.M.; Hoebink, J.H.B.J.; Marin, G.B.M.M.; Kruse, N.

    1998-01-01

    The oxidation of ethene by oxygen in the presence of steam and carbon dioxide over a commercially available Pt/Rh/CeO2/¿-Al2O3 three-way automotive catalyst was studied. Experiments were carried out in a fixed-bed micro reactor under intrinsic conditions, i.e. in the absence of external and internal

  13. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  14. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  15. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  16. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  17. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  18. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    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.

  19. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  20. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  1. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  2. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  3. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  4. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

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

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

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

  6. Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

  7. Performance evaluation of LaBr3: Ce scintillator

    International Nuclear Information System (INIS)

    Xie Ming; Lin Li; Liu Shihao; Xiao Peng; Xie Qingguo

    2012-01-01

    The cerium doped lanthanum bromide crystal (LaBr 3 : Ce) is a new kind of scintillator with many advantages such as good energy resolution, high light output, short decay time, good proportionality response. These properties make the LaBr 3 : Ce attractive substantial interest to use in the radiation detection. The energy resolution were investigated with Φ25 × 25 mm LaBr 3 : Ce coupled to a Hamamatsu R8900 photomultiplier tube. Energy resolution of 3.6% (FWHM) have been achieved for 511 keV photons ( 18 F source) at room temperature. Decay time constant of 20 ns have been acquired with a Hamamatsu fast-time-response R9800 photomultiplier tube. The results approve the excellent characterizations of LaBr 3 : Ce and imply its enormous potentiality in the radiation detectors of gamma-ray spectroscopy and PET. (authors)

  8. NMR study of CeCoSi3

    International Nuclear Information System (INIS)

    Iwamoto, Y.

    1995-01-01

    Low-temperature susceptibility, NMR and NQR of the 59 Co signal in CeCoSi 3 have been measured. CeCoSi 3 showed a superconducting transition at 0.7-1.2K. From NQR measurement, the nuclear quadrupole frequency and the full width at half maximum (FWHM) of 59 Co in CeCoSi 3 were estimated to be about 1.08MHz and 0.08MHz, respectively. The 59 Co nuclear spin-lattice relaxation rate (1/T 1 ) in CeCoSi 3 was proportional to the temperature (T) as the Fermi liquid state above the superconducting transition temperature (T c ), and then rapidly decreased below T c . ((orig.))

  9. CE-MS fingerprinting of Laurencia complex algae (Rhodophyta).

    Science.gov (United States)

    Machín-Sánchez, María; Asensio-Ramos, María; Hernández-Borges, Javier; Gil-Rodríguez, María Candelaria

    2014-03-01

    The use of CE-ESI-MS has been considered as a new chemical strategy for the possible discernment of genera and species of the Laurencia complex. After the selection of the CE-MS and the extraction conditions, a total of 28 specimens of the complex, including different species of four genera (Laurencia, Laurenciella, Palisada, and Osmundea) collected from five intertidal locations on the Island of Tenerife (Canary Islands, Spain) were analyzed. CE-MS fingerprints revealed that CE-MS can be used as a useful tool for these studies in order to assess similarities and differences between them and that it constitutes an important starting point for further studies in the field. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  10. Synthesis and Luminescent Characteristics of Ce3+-Activated Borosilicate Blue-Emitting Phosphors for LEDs

    Directory of Open Access Journals (Sweden)

    Hong Yu

    2016-01-01

    Full Text Available The phosphors Sr3B2SiO8:Ce3+ have been successfully synthesized via solid-state reaction process. Emission/excitation spectra and photoluminescence decay behaviors were investigated in detail. Under the excitation of 340 nm, the emission spectrum presented an asymmetry emission band extended from 350 to 600 nm, which with the main peak at 425 nm can be fitted in two peaks (23940 cm−1 and 21934 cm−1. The chromaticity coordinates of Sr3-xB2SiO8:xCe3+ are fixed in the blue region; when the intensity of Ce3+ reached the maximum, the chromaticity coordinate is (0.154, 0.088 which is more close to the standard CIE of blue light (0.140, 0.080. The results showed the kind of phosphor may have potential applications in the fields of UV-excited white LEDs.

  11. Uncertainty estimation of analysis of Fe, Ca, Zr, Ba, La, Ti and Ce in sediment sample using XRF method

    International Nuclear Information System (INIS)

    Sukirno; Agus Taftazani

    2010-01-01

    An uncertainty of analysis of Fe, Ca, Zr, Ba, La, Ti and Ce in river sediment of Panfuran Wariness sample by X RF method has been done. The result value of testing is meaningless if it isn't completed without uncertainty value. The calculation of Ba metal have been presented for example. The aim of the research is to get accreditation certificate of X-Ray Fluorescence method on laboratory of analytical PTAPB – BATAN as well as ISO guide 17025-2005. The result of calculation uncertainty of Fe, Zr, Ba, La, Ce, Ti and Ca analysis showed that the uncertainty components come from: preparation of sample and standard/comparator, purity of material, counting statistic (sample and standard ) and repeatability. The results showed that metals in river sediment of Pancuran Wonosari were Fe = 7.290%, Zr = 54.5 mg/kg, Ba = 1661.6 mg/kg, La = 22.9 mg/kg, Ce = 161.0 mg/kg, Ti = 3193.2 and Ca = 7.816%, and the result of uncertainty estimate of Fe, Zr, Ba, La, Ce, Ti and Ca were ± 0.60%, ± 4.5 mg/kg, ± 55 mg/kg, ± 1.4 mg/kg, 12.0 mg/kg, ± 208 mg/kg and ± 0.61%. (author)

  12. Reactor plant for Belene NPP completion

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

    2004-01-01

    Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

  13. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  14. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  15. Nitrate conversion and supercritical fluid extraction of UO2-CeO2 solid solution prepared by an electrolytic reduction-coprecipitation method

    International Nuclear Information System (INIS)

    Zhu, L.Y.; Duan, W.H.; Wen, M.F.; Xu, J.M.; Zhu, Y.J.

    2014-01-01

    A low-waste technology for the reprocessing of spent nuclear fuel (SNF) has been developed recently, which involves the conversion of actinide and lanthanide oxides with liquid N 2 O 4 into their nitrates followed by supercritical fluid extraction of the nitrates. The possibility of the reprocessing of SNF from high-temperature gas-cooled reactors (HTGRs) with nitrate conversion and supercritical fluid extraction is a current area of research in China. Here, a UO 2 -CeO 2 solid solution was prepared as a surrogate for a UO 2 -PuO 2 solid solution, and the recovery of U and Ce from the UO 2 -CeO 2 solid solution with liquid N 2 O 4 and supercritical CO 2 containing tri-n-butyl phosphate (TBP) was investigated. The UO 2 -CeO 2 solid solution prepared by electrolytic reduction-coprecipitation method had square plate microstructures. The solid solution after heat treatment was completely converted into nitrates with liquid N 2 O 4 . The XRD pattern of the nitrates was similar to that of UO 2 (NO 3 ) 2 . 3H 2 O. After 120 min of online extraction at 25 MPa and 50 , 99.98% of the U and 98.74% of the Ce were recovered from the nitrates with supercritical CO 2 containing TBP. The results suggest a promising potential technology for the reprocessing of SNF from HTGRs. (orig.)

  16. Thermal Expansion Property of U-Zr Alloys and U-Zr-Ce Alloys as a Surrogate Metallic Fuel for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Lee, Jong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Youn Myung; Lee, Chan Bock

    2010-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. An extensive database on the performance of advanced metal fuels was generated as a result of the operation of these reactors and the IFR program. In this study, the U-Zr binary alloys and U-Zr-Ce ternary alloys as surrogate metallic fuel were fabricated in lower pressure Ar environment by gravity casting. The melt temperature was approximately 1,500 .deg. C. Thermal expansion of the fuel during normal operation is related with fuel performance in a reactor. Therefore, it is necessary to investigate the thermal expansion of the fuel in order to warrant a good prediction the fuel performance

  17. Sr2CeO4: Electronic and structural properties

    International Nuclear Information System (INIS)

    Rocha, Leonardo A.; Schiavon, Marco A.; Nascimento, Clebio S.; Guimarães, Luciana; Góes, Márcio S.; Pires, Ana M.; Paiva-Santos, Carlos O.

    2014-01-01

    Highlights: • Sr 2 CeO 4 it was obtained from the heat treatment of Ce 3+ -doped strontium oxalate. • Rietveld analysis made it possible to obtain information about crystalline structure. • Experimental band gap value was compared with theoretical obtained by Sparkle/PM7. • The materials obtained shows intense photoluminescence and scintillator properties. - Abstract: This work presents on the preparation and photoluminescent properties of Sr 2 CeO 4 obtained from the heat treatment of Ce(III)-doped strontium oxalate (10, 25 and 33 mol%). The oxalate precursors were heat treated at 1100 °C for 12 h. The structure of this photoluminescent material was evaluated by the Rietveld method. The route used in this work to prepare the materials showed to be viable when compared to other synthesis reported in the literature. The Sr 2 CeO 4 material showed a broad and intense band emission with a maximum around 485 nm. The quantitative phase analysis showed that the Sr 2 CeO 4 photoluminescent phase is the majority one compared to the impurity phases of SrCeO 3 and SrCO 3 . From all results it was possible to verify a complete elimination of the CeO 2 phase for the sample obtained from the heat treatment of oxalate precursor containing 33 mol% of cerium(III). The material showed excellent properties for possible candidate as scintillator materials, and in the improvement of efficiency of solar cells when excited in the UV–vis region. The CIE chromaticity diagram it is also reported in this work

  18. Quadrupole moment of the superdeformed band in 131Ce

    International Nuclear Information System (INIS)

    He, Y.; Godfrey, M.J.; Jenkins, I.; Kirwan, A.J.; Nolan, P.J.

    1990-01-01

    A mean lifetime measurement has been carried out on the states in the superdeformed band found in 131 Ce using the Doppler shift attenuation method (DSAM). The measured intrinsic nuclear quadrupole moment is Q o approx= 6 eb, assuming constant deformation, which corresponds to a quadrupole deformation β 2 approx= 0.35. This is considerably smaller than the value deduced for 132 Ce. (author)

  19. Optical properties of CeO 2 thin films

    Indian Academy of Sciences (India)

    Cerium oxide (CeO2) thin films have been prepared by electron beam evaporation technique onto glass substrate at a pressure of about 6 × 10-6 Torr. The thickness of CeO2 films ranges from 140–180 nm. The optical properties of cerium oxide films are studied in the wavelength range of 200–850 nm. The film is highly ...

  20. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  1. Crystal growth and magnetic properties of equiatomic CeAl

    Science.gov (United States)

    Das, Pranab Kumar; Thamizhavel, A.

    2015-03-01

    Single crystal of CeAl has been grown by flux method using Ce-Al self-flux. Several needle like single crystals were obtained and the length of the needle corresponds to the [001] crystallographic direction. Powder x-ray diffraction revealed that CeAl crystallizes in orthorhombic CrB-type structure with space group Cmcm (no. 63). The magnetic properties have been investigated by means of magnetic susceptibility, isothermal magnetization, electrical transport, and heat capacity measurements. CeAl is found to order antiferromagnetically with a Neel temperature TN = 10 K. The magnetization data below the ordering temperature reveals two metamagentic transitions for fields less than 20 kOe. From the inverse magnetic susceptibility an effective moment of 2.66 μB/Ce has been estimated, which indicates that Ce is in its trivalent state. Electrical resistivity data clearly shows a sharp drop at 10 K due to the reduction of spin disorder scattering of conduction electrons thus confirming the magnetic ordering. The estimated residual resistivity ratio (RRR) is 33, thus indicating a good quality of the single crystal. The bulk nature of the magnetic ordering is also confirmed by heat capacity data. From the Schottky anomaly of the heat capacity we have estimated the crystal field level splitting energies of the (2J + 1) degenerate ground state as 25 K and 175 K respectively for the fist and second excited states.

  2. CE-SAM: a conversational interface for ISR mission support

    Science.gov (United States)

    Pizzocaro, Diego; Parizas, Christos; Preece, Alun; Braines, Dave; Mott, David; Bakdash, Jonathan Z.

    2013-05-01

    There is considerable interest in natural language conversational interfaces. These allow for complex user interactions with systems, such as fulfilling information requirements in dynamic environments, without requiring extensive training or a technical background (e.g. in formal query languages or schemas). To leverage the advantages of conversational interactions we propose CE-SAM (Controlled English Sensor Assignment to Missions), a system that guides users through refining and satisfying their information needs in the context of Intelligence, Surveillance, and Reconnaissance (ISR) operations. The rapidly-increasing availability of sensing assets and other information sources poses substantial challenges to effective ISR resource management. In a coalition context, the problem is even more complex, because assets may be "owned" by different partners. We show how CE-SAM allows a user to refine and relate their ISR information needs to pre-existing concepts in an ISR knowledge base, via conversational interaction implemented on a tablet device. The knowledge base is represented using Controlled English (CE) - a form of controlled natural language that is both human-readable and machine processable (i.e. can be used to implement automated reasoning). Users interact with the CE-SAM conversational interface using natural language, which the system converts to CE for feeding-back to the user for confirmation (e.g. to reduce misunderstanding). We show that this process not only allows users to access the assets that can support their mission needs, but also assists them in extending the CE knowledge base with new concepts.

  3. Chemical surveillance of commercial fast breeder reactors

    International Nuclear Information System (INIS)

    Stamm, H.H.; Stade, K.Ch.

    1988-01-01

    After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

  4. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  7. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  8. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  9. Cerocene Revisited: The Electronic Structure of and Interconversion Between Ce2(C8H8)3 and Ce(C8H8)2

    Energy Technology Data Exchange (ETDEWEB)

    Walter, Marc D.; Booth, Corwin H.; Lukens, Wayne W.; Andersen, Richard A.

    2009-02-02

    New synthetic procedures for the preparation of Ce(cot)2, cerocene, from [Li(thf)4][Ce(cot)2], and Ce2(cot)3 in high yield and purity are reported. Heating solid Ce(cot)2 yields Ce2(cot)3 and COT while heating Ce2(cot)3 with an excess of COT in C6D6 to 65oC over four months yields Ce(cot)2. The solid state magnetic susceptibility of these three organocerium compounds shows that Ce(cot)2 behaves as a TIP (temperature independent paramagnet) over the temperature range of 5-300 K, while that of Ce2(cot)3 shows that the spin carriers are antiferromagnetically coupled below 10 K; above 10 K, the individual spins are uncorrelated, and [Ce(cot)2]- behaves as an isolated f1 paramagnet. The EPR at 1.5K for Ce2(cot)3 and [Ce(cot)2]- have ground state of MJ= +- 1/2. The LIII edge XANES of Ce(cot)2 (Booth, C.H.; Walter, M.D.; Daniel, M.; Lukens, W.W., Andersen, R.A., Phys. Rev. Lett. 2005, 95, 267202) and 2Ce2(cot)3 over 30-500 K are reported; the Ce(cot)2 XANES spectra show Ce(III) and Ce(IV) signatures up to a temperature of approximately 500 K, whereupon the Ce(IV) signature disappears, consistent with the thermal behavior observed in the melting experiment. The EXAFS of Ce(cot)2 and Ce2(cot)3 are reported at 30 K; the agreement between the molecular parameters for Ce(cot)2 derived from EXAFS and single crystal X-ray diffraction data are excellent. In the case of Ce2(cot)3 no X-ray diffraction data are known to exist, but the EXAFS are consistent with a"triple-decker" sandwich structure. A molecular rationalization is presented for the electronic structure of cerocene having a multiconfiguration ground state that is an admixture of the two configurations Ce(III, 4f1)(cot1.5-)2 and Ce(IV, 4f0)(cot2-)2; the multiconfigurational ground state has profound effects on the magnetic properties and on the nature of the chemical bond in cerocene and, perhaps, other molecules.

  10. Comparison between the Oxygen Reduction Reaction Activity of Pd5Ce and Pt5Ce

    DEFF Research Database (Denmark)

    Tripkovic, Vladimir; Zheng, Jian; Rizzi, Gian Andrea

    2015-01-01

    A set of electrochemical and X-ray spectroscopy measurements have been used conjointly with density functional theory (DFT) simulations to study the activity and stability of Pd5Ce for the oxygen reduction reaction. A polycrystalline Pd5Ce rod has been selected as a model catalyst to test if resu......-Pd5Ce is more facile, requires less atom rearrangement, than transformation from Pt5Ce to Pt3Ce, which might explain the kinetic stability of Pt5Ce at low temperatures....

  11. Operator licensing examiner standards

    International Nuclear Information System (INIS)

    1994-06-01

    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

  12. Business Opportunities for Small Reactors

    International Nuclear Information System (INIS)

    Minato, Akio; Nishimura, Satoshi; Brown, Neil W.

    2007-01-01

    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)

  13. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    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. The modification of the Rossendorf Research Reactor

    International Nuclear Information System (INIS)

    Gehre, G.; Hieronymus, W.; Kampf, T.; Ringel, V.; Robbander, W.

    1990-01-01

    The Rossendorf Research Reactor is of the WWR-SM type. It is a heterogeneous water moderated and cooled tank reactor with a thermal power of 10 MW, which was in operation from 1957 to 1986. It was shut down in 1987 for comprehensive modifications to increase its safety and to improve the efficiency of irradiation and experimentals. The modifications will be implemented in two steps. The first one to be finished in 1989 comprises: 1) the replacement of the reactor tank and its components, the reactor cooling system, the ventilation system and the electric power installation; 2) the construction of a new reactor control room and of filtering equipment; 3) the renewal of process instrumentation and control engineering equipment for reactor operation, equipment for radiation protection monitoring, and reactor operation and safety documentation. The second step, to be implemented in the nineties, is to comprise: 1) the enlargement of the capacity for storage of spent fuel; 2) the modernization of reactor operations by computer-aided control; 3) the installation of an automated measuring systems for accident and environmental monitoring. Two objects of the modification, the replacement of the reactor tank and the design of a new and safer one as well as the increase of the redundancy of the core emergency cooling system are described in detail. For the tank replacement the exposure data are also given. Furthermore, the licensing procedures based on national ordinances and standards as well as on international standards and recommendations and the mutual responsibilities and activities of the licensing authority and of the reactor manager are presented. Finally, the present state of the modifications and the schedule up to the reactor recommissioning and test operation at full power is outlined

  15. System 80+{trademark} standard design incorporates radiation protection lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Crom, T.D.; Naugle, C.L. [Duke Engineering & Services, Inc., Charlotte, NC (United States); Turk, R.S. [ABB Combustion Engineering Nuclear Power, Windsor, CT (United States)

    1995-03-01

    Many lessons have been learned from the current generation of nuclear plants in the area of radiation protection. The following paper will outline how the lessons learned have been incorporated into the design and operational philosophy of the System 80+{trademark} Standard Design currently under development by ABB Combustion Engineering (ABB-CE) with support from Duke Engineering and Services, Inc. and Stone and Webster Engineering Corporation in the Balance-of-Plant design. The System 80+{trademark} Standard Design is a complete nuclear power plant for national and international markets, designed in direct response to utility needs for the 1990`s, and scheduled for Nuclear Regulatory Commission (NRC) Design Certification under the new standardization rule (10 CFR Part 52). System 80+{trademark} is a natural extension of System 80{sup R} technology, an evolutionary change based on proven Nuclear Steam Supply System (NSSS) in operation at Palo Verde in Arizona and under construction at Yonggwang in the Republic of Korea. The System 80+{trademark} Containment and much of the Balance of Plant design is based upon Duke Power Company`s Cherokee Plant, which was partially constructed in the late 1970`s, but, was later canceled (due to rapid declined in electrical load growth). The System 80+{trademark} Standard Design meets the requirements given in the Electric Power Research Institute (EPRI) Advanced Light Water Reactor (ALWR) Requirements Document. One of these requirements is to limit the occupational exposure to 100 person-rem/yr. This paper illustrates how this goal can be achieved through the incorporation of lessons learned, innovative design, and the implementation of a common sense approach to operation and maintenances practices.

  16. Determination of the O/M ratios of polynary uranium oxides by Ce(IV)-Fe(II) back titration after dissolution in mixed sulphuric and phosphoric acids

    International Nuclear Information System (INIS)

    Fujino, T.; Sato, N.; Yamada, K.

    1996-01-01

    Uranium (IV) in polynary uranium oxides is determined after the solid has been dissolved in a warm mixed solution of sulphuric and phosphoric acids containing excess Ce(IV). The latter is titrated with a Fe(II) standard solution using ferroin as indicator. This method is especially effective for (mixed) uranium oxides which are difficult to dissolve in hot Ce(IV) sulphuric acid. The standard deviation of the determined x value in polynary oxides is estimated to be below ± 0.004 for samples of 10-30 mg. (orig.)

  17. Determination of the O/M ratios of polynary uranium oxides by Ce(IV)-Fe(II) back titration after dissolution in mixed sulphuric and phosphoric acids.

    Science.gov (United States)

    Fujino, T; Sato, N; Yamada, K

    1996-01-01

    Uranium (IV) in polynary uranium oxides is determined after the solid has been dissolved in a warm mixed solution of sulphuric and phosphoric acids containing excess Ce(IV). The latter is titrated with a Fe(II) standard solution using ferroin as indicator. This method is especially effective for (mixed) uranium oxides which are difficult to dissolve in hot Ce(IV) sulphuric acid. The standard deviation of the determined x value in polynary oxides is estimated to be below +/- 0.004 for samples of 10-30 mg.

  18. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  19. The CE3R Network: current status and future perspectives

    Science.gov (United States)

    Lenhardt, Wolfgang; Pesaresi, Damiano; Živčić, Mladen; Costa, Giovanni; Kuk, Kresimir; Bondár, István; Duni, Llambro; Spacek, Petr

    2016-04-01

    In order to improve the monitoring of seismic activities in the border regions and to enhance the collaboration between countries and seismological institutions in Central Europe, the Environment Agency of the Slovenian Republic (ARSO), the Italian National Institute for Oceanography and Experimental Geophysics (OGS), the University of Trieste (UniTS) and the Austrian Central Institute for Meteorology and Geodynamics (ZAMG) established in 2001 the "South Eastern Alps Transfrontier Seismological Network". In May 2014 ARSO, OGS, UniTS and ZAMG agreed to formalize the transfrontier network, to name it "Central and East European Earthquake Research Network", (CE3RN or CE3R Network) in order to locate it geographically since cross-border networks can be established in other areas of the world and to expand their cooperation, including institutions in other countries. The University of Zagreb (UniZG) joined CE3RN in October 2014. The Kövesligethy Radó Seismological Observatory (KRSZO) of the Hungarian Academy of Sciences joined CE3RN in October 2015. The Institute of Geosciences, Energy, Water and Environment (IGEWE) of the Polytechnic University of Tirana joined CE3RN in November 2015. The Institute of Physics of the Earth (IPE) of the Masaryk University in Brno joined CE3RN in November 2015. CE3RN Parties intend to formalize and possibly extend their ongoing cooperation in the field of seismological data acquisition, exchange and use for seismological and earthquake engineering and civil protection purposes. The purpose of this cooperation is to retain and expand the existing cross-border network, specify the rules of conduct in the network management, improvements, extensions and enlargements, enhance seismological research in the region, and support civil protection activities. Since the formal establishment of CE3RN, several common projects have been completed, like the SeismoSAT project for the seismic data center connection over satellite funded by the Interreg

  20. The role of calcification for staging cystic echinococcosis (CE)

    Energy Technology Data Exchange (ETDEWEB)

    Hosch, Waldemar; Kauffmann, Guenter W. [University Hospital Heidelberg, Department of Radiology, Heidelberg (Germany); Stojkovic, Marija; Junghanss, Thomas [University Hospital of Heidelberg, Section of Clinical Tropical Medicine, Heidelberg (Germany); Jaenisch, Thomas [University Hospital of Heidelberg, Section of Clinical Tropical Medicine, Heidelberg (Germany); University Hospital of Heidelberg, Section of Biostatistics and Epidemiology, Heidelberg (Germany)

    2007-10-15

    The prevalence of calcified cysts and the significance of calcification as a sign of cyst inactivity in cystic echinococcosis (CE) was evaluated. Seventy-eight patients (36 females, 42 males, mean age 40.8 {+-} 16.9 years) with CE, having a total of 137 abdominal cysts (116 hepatic, three splenic, one renal and 17 peritoneal cysts), were diagnosed and followed-up by ultrasound during and after albendazole treatment or as part of the watch-and-wait approach recording changes in the cyst wall and content. In 48 patients with 94 cysts, computed tomography (CT) imaging was additionally available and was correlated with ultrasound findings. Cyst wall calcification was classified into (1) ''sprinkled'', (2) ''eggshell-like'', and (3) ''circular''. Calcification of the cyst wall and/or cyst content was detected in 67 echinococcal cysts (48.9% of all cysts) in 39 patients (15 females, 24 males, mean age 40.8 {+-} 14.8 years). Of the total of 67 calcified cysts, only 23 were compatible with WHO type CE5, 18 with WHO type CE4. Judged by cyst content, the remaining 26 were of WHO type CE1, CE2 and CE3 (n = 1, n = 8, and n = 17, respectively). During a mean period of 34.3 months ({+-}21.3 months) the majority of cysts (n = 32) did not exhibit any change in cyst content and wall properties. Fourteen cysts showed signs of progressive involution, five cysts (all of WHO type CE3) of renewed activity defined by recurring fluid collection. In 16 cysts, no follow-up was available due to surgery or drop out. Calcification of the cyst is not restricted to the inactive WHO cyst types CE4 and CE5, but occurs in all stages and in up to 50% of cysts. The completeness and, most importantly, the stability of consolidation of cyst content over time predicts cyst inactivity more reliably. (orig.)