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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. 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

  3. Inspection technology of Korean standard Reactor vessel

    International Nuclear Information System (INIS)

    Doh, Euisoon

    2002-01-01

    This paper is for the inspection technology of Korean standard Reactor vessel. KPS performed Korean standard Reactor vessel inspection for a part of Yonggwang unit 5 and full inspection of Yonggwang unit 6 on November in 2001. Korean standard Reactor vessel is different from Westinghouse and Framatome plants in design and material. Understanding design, structure and material is very important to prepare and to perform Reactor vessel inspection. Korean standard Reactor vessel has 6 Stabilizer lugs, 6 Core stop lugs and one Flow Baffle around Lower head to Bottom head weld(G4). They make very difficult to perform scanning of G4 by transducers and to install Lower Platform of robot system. Korean standard Reactor vessel has 2 Outlet Nozzles, 4 Inlet Nozzles and small curvature of Inner Radius regions. That small curvature's Inner Radius region needs a special technology to perform the area more than 90% of whole examination volume. The most significant difference of Korean standard Reactor vessel is that it does not have Safe End region because both of Reactor vessel and main reactor coolant system piping are carbon steel. Therefore, the technology for this nozzle to piping weld is very unique to be compared with other Westinghouse or Framatome plants. Described above, the inspection technology of Korean standard Reactor Vessel is very different from other Reactor Vessel due to different internal structure and material. The company which performs this 10 years ISI in future will need much investigation of structure and material to succeed in Reactor vessel inspection of Korean standard plant

  4. 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

  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. 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

  7. 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

  8. 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.

  9. 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)

  10. 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

  11. ABB-CE2s System 80+ advanced light water reactor design

    International Nuclear Information System (INIS)

    Flynn, E. P.

    1994-01-01

    This presentation will focus on the evolutionary improvements that have been made to System 80 which is the current operating design that was selected as the basis for developing System 80+. The Electric Power Research Institute (EPRI) has developed and issued the Advanced Light Water Reactor (ALWR) requirements as the basis for future ALWR's. ABB-CE and its associates were active participants in the development of the ALWR design requirements. System 80+ development has been in accordance with the EPRI requirements and System 80+ is the only Evolutionary Pressurized Water Reactor design to these requirements. The System 80+ design currently is in the final stages of NRC review leading to certification as a complete power plant1 Nuclear Island, Turbine lsland, and Balance of Plant, which are integrated for safe and reliable operation. Improved safety is a major feature of System 80+ with redundancy and diversity providing a prudent balance between accident prevention and mitigation. Safety Injection and Emergency Feedwater are dedicated four-train systems. The containment structure is a large spherical steel vessel surrounded by a cylindrical concrete shield building. The entire nuclear Island is founded on a common basemat to provide seismic resistance. The Nuclear Steam Supply System design has incorporated increased margin and improved reliability. High technology focused the control complex on the human operator. Programmable logic controllers, distributed system architecture, super mini-computers, fiber optic communications, touch sensitive cathode ray tube and advanced signal validation tecniques greatly simplify the design

  12. 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

  13. 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

  14. 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

  15. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-22

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

  16. Reference design for the standard mirror hybrid reactor

    International Nuclear Information System (INIS)

    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

  17. 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

  18. 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

  19. 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

  20. 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

  1. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    Energy Technology Data Exchange (ETDEWEB)

    GRIFFIN, PATRICK J.

    1999-09-14

    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.

  2. 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)

  3. 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

  4. 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)

  5. 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

  6. 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

  7. 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)

  8. Efficient degradation of chlorobenzene in a non-thermal plasma catalytic reactor supported on CeO2/HZSM-5 catalysts.

    Science.gov (United States)

    Jiang, Liying; Nie, Guofeng; Zhu, Runye; Wang, Jiade; Chen, Jianmeng; Mao, Yubo; Cheng, Zhuowei; Anderson, Willam A

    2017-05-01

    Chlorobenzene removal was investigated in a non-thermal plasma reactor using CeO 2 /HZSM-5 catalysts. The performance of catalysts was evaluated in terms of removal and energy efficiency. The decomposition products of chlorobenzene were analyzed. The results show that CeO 2 /HZSM-5 exhibited a good catalytic activity, which resulted in enhancements of chlorobenzene removal, energy efficiency, and the formation of lower amounts of by-products. With regards to CO 2 selectivity, the presence of catalysts favors the oxidation of by-products, leading to a higher CO 2 selectivity. With respect to ozone, which is considered as an unavoidable by-product in air plasma reactors, a noticeable decrease in its concentration was observed in the presence of catalysts. Furthermore, the stability of the catalyst was investigated by analyzing the evolution of conversion in time. The experiment results indicated that CeO 2 /HZSM-5 catalysts have excellent stability: chlorobenzene conversion only decreased from 78% to 60% after 75hr, which means that the CeO 2 /HZSM-5 suffered a slight deactivation. Some organic compounds and chlorinated intermediates were adsorbed or deposited on the catalysts surface as shown by the results of Fourier Transform Infrared (FT-IR) spectroscopy, scanning electron microscope (SEM) and energy dispersive X-ray spectroscopy (EDS) analyses of the catalyst before and after the reaction, revealing the cause of catalyst deactivation. Copyright © 2016. Published by Elsevier B.V.

  9. 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...

  10. American National Standard: fire-protection program criteria for research reactors

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    This standard provides criteria for a fire protection program for research reactor facilities and for the reactor safety-related systems included in those facilities. It stresses preservation of the capability to achieve and maintain safe shutdown of the reactor, and includes consideration of both direct fire hazards and indirect or consequential hazards

  11. 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

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

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

  13. 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

  14. 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 ...

  15. Reactor physics standards: The key to successfully dealing with technical diversity

    International Nuclear Information System (INIS)

    Knuckles, E.R.

    1990-01-01

    Effective and valuable reactor physics standards can be successfully developed to accommodate diversity in available calculations tools and promote improvement in existing methods. The issues encountered and lessons learned in the standard, 'Calculation of Doppler Reactivity for Use in Thermal Light Water Reactor Analysis' (now under development by the ANS 19.7 working group), demonstrate this point. With a diversity of reactor physics tools available, differing levels of user experience, and a variety of procedures for calculating reactor physics parameters important to safety, it is not surprising that there are differing levels of quality in the calculational result of the same parameters. An approach that effectively deals with this technical diversity is standardization of the calculational process. This approach assures that the user's expectations are consistently met. In order for a standard to be effective, it must recognize and address three essential elements of this process: the user, a set of codes, and associated procedures

  16. Conversion and standardization of US university reactor fuels using LEU, status 1989

    International Nuclear Information System (INIS)

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

    1989-01-01

    In 1986, the US Department of Energy initiated a program to change the fuel used in most of the US university research reactors using HEU (93%) to LEU(<20%) in order to minimize the risk of theft or diversion of this weapons-useable material. An important consideration in the LEU conversion planning process has been the desire to standardize the fuels that are used and to enhance the performance and utilization of the reactors. This paper describes the current status of this conversion process and the plans and schedules to complete an orderly transition from HEU to LEU fuel in most of these reactors. To date, three university reactors have been converted to LEU fuel, completed safety documentation for three reactors is being evaluated by the USNRC, and work on the safety documentation for six reactors is in progress. 13 refs., 9 tabs

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. 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 effect of the infiltration on the electrochemical properties and catalytic activity of the reactor was investigated by electrochemical impedance spectroscopy (EIS) and gas analysis. Figure 1 shows the SEM cross-section micrograph of the electrochemical reactor made of 11 alternating layers of electrode...... matter, which lead to the formation of ozone in urban and regional areas [1]. The electrocatalytic activity of a porous electrochemical reactor, made of La0.85Sr0.15MnO3±δ (LSM) as electrode and Ce0.9Gd0.1O1.95 (GDC) as electrolyte, was studied for the electrochemical oxidation of propene (C3H6), a major...... (LSM) and electrolyte (GDC). Figure 2 shows the Nyquist plot of the impedance spectra of Co3O4 infiltrated backbone recorded at OCP with 10% O2 and 10% O2 + 1000 ppm C3H6, 2 L/h, 400 °C. [1] R. Atkinson, Atmospheric Chemistry of VOCs and NOx, Atmos Environ. 34 (2000) 2063; [Formula]...

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. Implementation of USP antibody standard for system suitability in capillary electrophoresis sodium dodecyl sulfate (CE-SDS) for release and stability methods.

    Science.gov (United States)

    Esterman, Abbie L; Katiyar, Amit; Krishnamurthy, Girija

    2016-09-05

    Capillary electrophoresis sodium dodecyl sulfate (CE-SDS) is widely used for purity analysis of monoclonal antibody therapeutics for release and stability to demonstrate product consistency and shelf life during the manufacturing and life cycle of the product. CE-SDS method development is focused on exploring the method capability to provide the information about the product purity and product related degradants (fragmentation, aggregation etc.). In order to establish the functionality of the instrumentation, software, and sample preparation; system suitability criteria need to be defined for analytical methods using a well characterized reference standard run under the same protocol and analysis as the test articles. Typically the reference standard is produced using a manufacturing process representative of the clinical material. The qualification, control, and maintenance of in-house reference standards are established through rigorous quality and regulatory guidelines. The U.S. Pharmacopeia (USP) has developed a monoclonal IgG System Suitability Reference Standard to be utilized for assessment of system suitability in CE-SDS methods. In this communication, we evaluate the system suitability acceptance criteria performance of the USP IgG standard using two methods, the recommended USP protocol provided in monograph and a molecule specific Bristol-Myers Squibb (BMS) CE-SDS method. The results from USP IgG standard were compared with two in-house monoclonal antibody reference standards. The data suggest that the USP CE-SDS method may not be suitable for CE-SDS analysis for release and stability of monoclonal antibody therapeutics due to the high level of method induced partial reduction observed for all molecules tested. This high level of fragmentation observed utilizing the USP method will result in reporting lower purity levels, which will impact the overall quality assessment of the molecule. The system suitability criteria recommended by the USP method can be

  4. 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

  5. Impedance Spectroscopy and Catalytic Activity Characterization of a La0.85Sr0.15MnO3/Ce0.9Gd0.1O1.95 Electrochemical Reactor for the Oxidation of Propene

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    This study aims to characterize the catalytic and electrochemical behavior of a La0.85Sr0.15MnO3/Ce0.9Gd0.1O1.95 porous reactor for the oxidation of propene in the presence of oxygen. The application of anodic polarization strongly increased the propene oxidation rate up to 71 %, although...... the current efficiency remained low. The effect of prolonged polarization on the reactor catalytic activity was evaluated. Prolonged polarization enhanced both the reactor intrinsic catalytic activity and the electrode performance due to the formation of oxygen vacancies on the electrode surface...

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

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

    1994-08-01

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

  7. 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)

  8. 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

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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...

  14. 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

  15. American National Standard: design requirements for light-water-reactor fuel-handling systems

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    This standard defines the required functions of fuel handling systems at light water reactor nuclear power plants. It provides minimum design requirements for equipment and tools for safe handling of nuclear fuel and control components at light water reactor nuclear power plants. The fuel handling system covered by this standard consists of handling equipment used for receiving and inspecting fuel containing new and recycled uranium; transporting on-site and loading fuel containing new and recycled uranium or irradiated fuel and control components in the reactor; removing from the reactor, transporting to storage, and inspecting irradiated fuel and loading casks for shipment of irradiated fuel from the site. It includes basic requirements and configuration for design, fabrication, maintenance, and operation. The basis of this standard is that the intended function of the equipment will be performed in an efficient and economical manner that assures protection to plant personnel and to the public, and that any radiation exposurers are maintained as low as reasonably achievable

  16. 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

  17. 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.)

  18. 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.

  19. 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

  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. 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)

  2. 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.)

  3. Standard Practice for Determining NeutronExposures for Nuclear Reactor Vessel Support Structures

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for: 1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities. 1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures. 1.2 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E > 1 MeV) that exceeds 1 × 1017 neutrons/cm2 or 3.0 × 10−4 dpa. (See Terminology E 170.) 1.3 Exposure of vessel support structures by gamma radiation is not included in the scope of this practice, but see the brief discussion of this issue in 3.2. 1.4 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 h...

  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 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)

  6. 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

  7. R and D issues in structural design standard for commercialized fast reactor components

    International Nuclear Information System (INIS)

    Shibamoto, Hiroshi; Tanaka, Yoshihiko; Inoue, Kazuhiko; Kasahara, Naoto; Morishita, Masaki

    2003-01-01

    Conceptual design studies of Japanese commercialized fast reactors (FRs) are carried out. With careful considerations on safety, economical improvement for practical use of these plants are aimed for, and the design of them is rationalized with adopting simple and innovative components, etc.. To certify the design concepts and validate structural integrity, research and development of Fast Reactor Structural Design Standard (FDS) for commercialized fast reactor components is now under way. Based on general characteristics of FRs and design needs of commercialized FRs, main subjects of R and D were identified. As for failure criteria, development for rational treatment of creep design region is dealt with. Ratcheting fatigue tests are conducted for the purpose of confirming the limit that ratcheting strain has negligible effects on structure strength. Besides, objecting to contribute to design rationalization through prediction of elasto-plastic and creep behaviors with high-precision, efforts to establish inelastic analysis methodologies for design are carried out. A guideline on inelastic analysis for design related to FDS, is prepared. This guideline is applied to evaluate the structural integrity of critical parts of components in conceptual design of commercialized FRs. Furthermore, aimed for mitigating thermal loads, a guideline on thermal loads modeling for design related to FDS is under developing. (author)

  8. 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

  9. 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...

  10. 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.

  11. 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)

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

    International Nuclear Information System (INIS)

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

    2003-02-01

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

  13. Toward Global Standards on Peaceful uses of Space Nuclear Reactor Power Systems

    Science.gov (United States)

    El-Genk, M. S.

    Space reactor power systems are enabling to future space exploration and outposts on the Moon, Mars and other celestial bodies. They could provide for electrical propulsion, thermal propulsion and bimodal operation of thermal propulsion and electrical power generation. The much higher specific impulse of nuclear thermal and electric propulsion, compared to chemical rockets, could significantly shorten the travel time to Mars and farthest destinations, reducing the exposure of equipment and astronauts to the harmful space radiation. This article addresses safety issues relevant to the design, operation and end-of-life storage of these systems in an attempt to stimulate a discussion of the need to establish specific and globally agreed upon safety standards and guidelines for future peaceful uses.

  14. 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.

  15. 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

  16. 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...

  17. 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.)

  18. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is a neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch. 1 claim, 16 figures

  19. Ce-Zr-La/Al2O3 prepared in a continuous stirred-tank reactor: a highly thermostable support for an efficient Rh-based three-way catalyst.

    Science.gov (United States)

    Wang, Su-Ning; Lan, Li; Hua, Wei-Bo; Shi, Zhong-Hua; Chen, Yao-Qiang; Gong, Mao-Chu; Zhong, Lin

    2015-12-21

    Two Ce-Zr-La/Al2O3 composite oxides, CZLA-C and CZLA-B, were synthesized using a co-precipitation method in a continuous stirred-tank reactor (CSTR) and a batch reactor (BR), respectively. Two Rh-based three-way catalysts (TWCs), Rh/CZLA-C and Rh/CZLA-B were obtained by a wet-impregnation method using the two composites as the supports. The physicochemical properties of the samples before and after thermal treatment at 1000 °C were characterized by N2 adsorption-desorption, X-ray diffraction (XRD), scanning electron microscopy (SEM), transmission electron microscopy (TEM), high-resolution transmission electron microscopy (HRTEM), H2-temperature programmed reduction (H2-TPR) and CO chemisorption. The results indicated that CZLA-C shows higher thermal stability than CZLA-B due to a sparsely-agglomerated morphology. Compared with Rh/CZLA-B, Rh/CZLA-C displayed better reducibility and higher thermal stability and exhibited significantly higher activity in the catalytic removal of the simulated gasoline vehicle exhaust emission (NO, CO and C3H8). Our work can provide a facile and economical synthesis route to advanced support materials and catalysts for exhaust emission control.

  20. 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

  1. 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...

  2. 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.

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

    International Nuclear Information System (INIS)

    Benjelloun, M.

    1984-01-01

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

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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

  15. Design of a weapons-grade plutonium assembly for optimal burnup in a standard pressurized water reactor

    Science.gov (United States)

    Alonso-Vargas, Gustavo

    We created a new MOX fuel assembly design that can be used in standard Westinghouse pressurized water reactors (PWR) to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which is called MIX-33, appears to be a good option for the disposition of weapons-grade plutonium (WG-Pu), increasing the plutonium disposition rate by 8% compared to a previous Westinghouse design. It is based in two novel ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the increase of the moderation ratio of the assembly. We replaced 8 fuel pins by water holes to increase the moderation ratio. We can transition smoothly from a full LEU core to a full MIX-33 core meeting the operational and safety regulations of a standard PWR. Given a MOX supply interruption scenario we can transition smoothly to full LEU meeting the safety regulations and using standard LEU assemblies with uniform enriched pin-wise distribution. If the MOX supply is interrupted for only one cycle, we are able to transition back to full MIX-33 core. However, in this case we probably need to de-rate the power by a few percent for a few weeks at the beginning of the cycle (BOC) to accommodate high peaking. For comparison we created another assembly design without extra water holes, which we called "MIX-25". It behaves in all the conditions analyzed in a similar way to the MIX-33 but it does present minor control problems. These can be solved by making small modifications to the control and safety systems, namely by enriching the boron-10 content of some boron absorbers. Thus, the addition of water holes replacing fuel pins helps to improve the MIX-33 performance and eliminate the difficulties seen in the MIX-25 design. We also performed a benchmarking analysis to test the code CASMO-3 to analyze WG-Pu assemblies, using the code MCNP-4A to compare. We found good agreement between CASMO-3 and

  16. 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

  17. Bridging the gap: Research and validation of the DST (Dynamic System Testing) performance test method for CE and ISO standards. Progress and preliminary results

    NARCIS (Netherlands)

    Naron, D.J.; Ree, B.G.C. van der; Rolloos, M.

    1998-01-01

    The Dynamic System Testing (DST) method is one of the methods chosen in the preliminary CEN/TC 312 European quality standards [1 ] to measure the energetic performance of Solar Domestic Hot Water (SDHW) systems. These European standards make a reference to the DST procedure as defined in the Draft

  18. 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

  19. 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.)

  20. 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

  1. 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.

  2. 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.''

  3. Evaluation of system requirements and standards development for thermal annealing of reactor pressure vessels. Progress report, March-September 1982

    International Nuclear Information System (INIS)

    Server, W.L.

    1983-03-01

    The material property data on thermal annealing of reactor-pressure-vessel steels have been reviewed. The most-critical materials are high copper welds; the data indicate that close to full recovery of Charpy V-notch properties can be realized by annealing at 850 0 F for 1 week (168 hours). However, the variability and sparcity of annealing recovery data dictate appropriate surveillance and experimental test programs. Of particular concern are the actual fracture-toughness changes and the differences between test- and power-reactor conditions. A survey of pressurized-water-reactor vendors, architect engineers, and consultants has been performed to assess the feasibility of performing a thermal-annealing cycle to restore the beltline material properties to their original state. This survey also addressed the issue of whether or not an actual annealing demonstration should be performed. It is the consensus view of the industry that in-situ reactor-vessel annealing can be done, but several areas still need to be studied and the American Society of Mechanical Engineers' Boiler and Pressure Vessel Code guidelines need to be defined

  4. How Temelin is being brought in line with Western standards. [Modifications to design of VVER type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simerka, I. (Jaderna Elektrana, Temelin (Czechoslovakia))

    1991-03-01

    Following independence from the USSR, the two Sovient-designed VVER 1000 Pressurized Water Reactors under construction at Temelin in Czechoslovakia are being extensively modified. The recommendations of IAEA missions are being applied and the station staff are reappraising the original design to improve safety and performance. (author).

  5. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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...

  11. 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...

  12. 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...

  13. 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

  14. Seamless Ligation Cloning Extract (SLiCE) cloning method.

    Science.gov (United States)

    Zhang, Yongwei; Werling, Uwe; Edelmann, Winfried

    2014-01-01

    SLiCE (Seamless Ligation Cloning Extract) is a novel cloning method that utilizes easy to generate bacterial cell extracts to assemble multiple DNA fragments into recombinant DNA molecules in a single in vitro recombination reaction. SLiCE overcomes the sequence limitations of traditional cloning methods, facilitates seamless cloning by recombining short end homologies (15-52 bp) with or without flanking heterologous sequences and provides an effective strategy for directional subcloning of DNA fragments from bacterial artificial chromosomes or other sources. SLiCE is highly cost-effective and demonstrates the versatility as a number of standard laboratory bacterial strains can serve as sources for SLiCE extract. We established a DH10B-derived E. coli strain expressing an optimized λ prophage Red recombination system, termed PPY, which facilitates SLiCE with very high efficiencies.

  15. 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.

  16. 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

  17. 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)

  18. 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

  19. 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...), Section 14.3.12 on ``Physical Security Hardware--Inspections, Tests, Analyses, and Acceptance Criteria... Security and Incident Response (NSIR) is revising SRP Section 14.3.12 (Enclosure 1), which updates the...

  20. CE-BEMS

    DEFF Research Database (Denmark)

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

    2016-01-01

    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...... 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...

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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

  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. 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...

  9. Oxidation of carbon by CeO{sub 2}: Effect of the contact between carbon and catalyst particles

    Energy Technology Data Exchange (ETDEWEB)

    May Issa; Corinne Petit; Alain Brillard; Jean-Francois Brilhac [Universite de Haute-Alsace, Mulhouse (France). Laboratoire Gestion des Risques et Environnement

    2008-05-15

    The oxidation of carbon black, CB, in presence of CeO{sub 2} is investigated to gain a better understanding of the effect of the contact between the two solids during this reaction. Different CB/CeO{sub 2} mixtures are tested in a fixed bed reactor. The experimental data are used to propose a model for CB oxidation in presence of CeO{sub 2}. It accounts for the size distribution of CeO{sub 2} particles, the contact area between CB and CeO{sub 2}, the mass of CB in the sample and the initial ratio CB/CeO{sub 2}. Corresponding kinetic parameters are determined. 35 refs., 10 figs., 3 tabs.

  10. 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.

  11. 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

  12. 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.

  13. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  14. 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

  15. 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

  16. Novel nanostructured CeO2 as efficient catalyst for energy and ...

    Indian Academy of Sciences (India)

    ronment by combustion-based technologies. Automo- bile and industrial exhausts such as carcinogenic soot particles and poisonous CO molecules are major ..... For a typical CO oxidation activity measurement,. ∼40 mg of Pt/CeO2 sample was placed in a U-shaped quartz micro-reactor (internal diameter of 4 mm) and.

  17. 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)

  18. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-09-01

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

  19. 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.)

  20. Reactor Neutrinos

    OpenAIRE

    Kim, Soo-Bong; Lasserre, Thierry; Wang, Yifang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  1. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  2. Reactor vessel

    NARCIS (Netherlands)

    Makkee, M.; Kapteijn, F.; Moulijn, J.A.

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and

  3. 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)

  4. CER. Research reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, Jerome [CEA, DEN, DER, Saint-Paul-lez-Durance (France). Jules Horowitz Reactor (JHR)

    2012-10-15

    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.)

  5. 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.)

  6. 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

  7. 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)

  8. 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.

  9. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Nagatomi, Shozo.

    1976-01-01

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

  10. 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.

  11. 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

  12. Ce isotope systematics of island arc lavas from the Lesser Antilles

    Science.gov (United States)

    Bellot, Nina; Boyet, Maud; Doucelance, Régis; Pin, Christian; Chauvel, Catherine; Auclair, Delphine

    2015-11-01

    The La-Ce systematics has one of the longest half-lifes (T1/2 = 292.5 Ga) of radioactive decay systems used in isotope geochemistry. Variations of the 138Ce/142Ce ratio are expected to be small and the use of Ce as isotopic tracer requires a very precise measurement. Compared to Sm-Nd studies, the La-Ce decay system can provide additional information about the nature of sediments recycled in subduction zones, because unusually large Ce anomalies relative to the neighboring rare earth elements exist in marine sediments such as fish teeth or hydrothermal deposits. Here, we present a chemical purification technique for Ce, and mass spectrometric technique to perform accurate and reproducible analyses of Ce isotopes of natural samples. We report a large set of Ce isotope data including analysis of 2 Ce reference material solutions (AMES and JMC-304), 2 rock standards (BCR-2 and BHVO-2), 2 chondrites (the carbonaceous chondrite Allende and the enstatite chondrite Sahara 97072), 4 mid-ocean ridge basalts, 30 arc lavas from the Martinique Island and 5 oceanic sediments from DSDP-site 144 drilled on the Demerara rise. The long-term, external precision obtained on the AMES reference material is 80 ppm (2 s.d., 138Ce/142Ce = 0.0225732 ± 18, n = 89). However, we note an evolution of isotopic ratios measured in static mode over the duration of this study (33 months). When the reproducibility is calculated from the AMES reference material measured during the same analytical session, it averages 40 ppm. All the 138Ce/142Ce ratios have been normalized to the AMES value of 0.0225746 (measured in session 7, 2 s.d. = 14 ppm, n = 8), a session during which the chondritic value has been defined and the peak tailing was negligible. The 138Ce/142Ce ratio measured for the JMC-304 Ce reference reagent is 0.0225706 ± 9 (2 s.d. = 38 ppm, n = 10). The analytical precision on natural samples is improved by a factor of about 4 in relation to previous studies on island arcs (Tanaka et al

  13. SPRAY CALCINATION REACTOR

    Science.gov (United States)

    Johnson, B.M.

    1963-08-20

    selfdiffusion coefficient for lanthanum, cerium, and praseodymium were determined. The investigation of phase relationships in the plutonium-cerium-copper ternary system was continued on samples containing a high concentration of copper. These analyses indicate that complete solid solution exists between the binary compounds CeCu/sub 2/ and PuCu/sub 2/, thus forming a quasi-binary system. The study of high temperature ceramic fuel materials has continued with the homogenization and microspheroidization of binary mixtures of plutonium dioxide and zirconium dioxide. Sintering a die-pressed pellet of the mixed powders for one hour at 1450 deg C was not sufficient to completely react the constituents. Complete homogenization was obtained when the pellet was melted in the plasma flame. In addition to the plutonium dioxide-zirconium dioxide microspheres, pure beryllium oxide microspheres were produced in the plasma torch. The electronic distribution functions for the 10% by weight PuO/sub 2/ dissolved in a silicate glass were determined. The plutonium-oxygen interaction at about 2.2A is less than the plutonium-oxygen distance for the 5% PuO/sub 2/. The decrease in the interionic distance is indicative of a stronger plutonium-oxygen association for the more concentrated composition. Potassium plutonium sulfate is being evaluated as a reagent to quantitatively separate plutonium from aqueous solutions. The compound containing two waters of hydration was prepared for thermogravimetric studies using analytically pure plutonium-239. Because of the stability of this compound, it is being evaluated as a calorimetric standard for plutonium-238. (auth)

  14. 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

  15. 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

  16. 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.)

  17. 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

  18. 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)

  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. 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

  1. 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

  2. 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)

  3. CuO/CeO2 catalysts prepared with different cerium supports for CO oxidation at low temperature

    International Nuclear Information System (INIS)

    Lu, Chi-Yuan; Chang, Wen-Chi; Wey, Ming-Yen

    2013-01-01

    The activity of a catalyst depends on the nature of its support, its active site, and its preparation method. This study aimed to employ various types of CeO 2 supports such as commercial CeO 2 and self-prepared CeO 2 for the preparation of copper catalysts. The CuO/CeO 2 catalysts were prepared using the polyol process and impregnation method. The catalysts were characterized using Brunauer–Emmett–Teller analysis, scanning electron microscopy, and X-ray analysis, and their catalytic activity for CO removal was evaluated in a microcatalytic reactor. The experimental results showed that the catalytic activity of the CuO/CeO 2 catalysts with different calcination temperatures decreased in the following order: 500 °C > 300 °C > 700 °C. Compared to the impregnation method, the polyol process generated well-dispersed metal particles over the support and showed higher CO removal efficiency with low activation energy. Compared to CuO/CeO 2 catalysts with commercial CeO 2 , those with CeO 2 that was self-prepared by pyrolysis had a large pore volume and good crystal structure of CeO 2 and showed good performance. The catalytic activity for CO removal was in the following order: CuO/CeO 2 -P (pyrolysis) > CuO/CeO 2 -C (commercial) > CuO/CeO 2 -D (deposition precipitation). CuO/CeO 2 -P catalysts showed good activity even at low temperature. The CuO/CeO 2 -P(300)-P-120 min catalyst was found to possess the good CO removal rate when the oxygen content was 6%, CO concentration was 500 ppm, catalyst weighed 1.0 g, pollutant gas velocity was 500 mL min −1 , SV was 3.7 × 10 4 h −1 , and reaction temperature was 150 °C. - Highlights: • CuO/CeO 2 catalysts were prepared using polyol and impregnation methods. • The supports of catalyst were self-prepared cerium oxide and commercial cerium oxide. • Pyrolysis and deposition precipitation methods were used for cerium preparation. • Catalytic activity of CuO/CeO 2 catalysts were evaluated by the removal of CO

  4. 40 CFR 63.1406 - Reactor batch process vent provisions.

    Science.gov (United States)

    2010-07-01

    ... megagram of product or less for non-solvent-based resin production. (2) The owner or operator of a reactor... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor...

  5. SLOWPOKE reactor

    International Nuclear Information System (INIS)

    Evans, D.J.R.; Downs, W.E.

    1974-01-01

    The SLOWPOKE reactor is described, which is a small pool type with thermal neutron fluxes ranging from 10 11 -10 12 n cm -2 sec -1 . It differs in many ways from conventional pool type, namely small critical mass, beryllium reflector and a closed reactor container. The reactor is designed as small and simply as possible, and consistently with safety and good operating practice. Access to the present model is via pneumatic irradiation tubes only, which limits the use of the facility to activation analysis, tracer production and training. (Mori, K.)

  6. 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

  7. 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

    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...... concentration of stable oxygen species on the electrode surface....

  8. 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

  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. 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.

  11. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  12. 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.

  13. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

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

  14. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  15. 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.)

  16. Propulsion reactors

    International Nuclear Information System (INIS)

    Anon.

    1999-01-01

    A nuclear reactor equips the recently constructed French aircraft- carrier Charles-De-Gaulle, in a few months the second nuclear submarine (SNLE) of new generation will be operational. In last october the government launched the program Barracuda which consists of 6 submarines (SNA) whose series head will be operational in 2010. The main asset of nuclear propulsion is to allow an almost unlimited autonomy: soft water, air are produced inside the submarine and the maximum time spent underwater is only limited by human capacity to cope with confinement. CEA has 3 missions concerning country defence. First the designing, the fabrication and the maintenance of weapons, secondly the supplying of fissile materials and thirdly the nuclear propulsion. A new generation of propulsion reactors is being studied and a ground installation involving a test reactor equivalent to that on board is being built. This test reactor (RES) will simulate any type of on-board reactors by adjusting temperature, pressure, flowrate and even equipment such as steam generator. This reactor will validate the technological choices for the Barracuda program. (A.C.)

  17. 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

  18. 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.

  19. 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.

  20. Integrity of the reactor vessel support system for a postulated reactor vessel closure head drop event

    International Nuclear Information System (INIS)

    Kim, Tae Wan; Lee, Ki Young; Lee, Dae Hee; Kim, Kang Soo

    1996-01-01

    The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor closure head drop event. 9 refs., 10 figs., 3 tabs. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

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

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. 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)

  7. 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)

  8. 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

  9. 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.

  10. 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)

  11. Reactor utilization

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-01-01

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

  12. 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.

  13. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Araki, Takao; Saito, Yasushi.

    1987-01-01

    Purpose: To reduce the seismic wave responsivity of an exhaust duct shields thereby preventing the release of tritium in an evacuating device due to failures upon earthquakes. Constitution: The ends on the cutting side of upper outer exhaust duct shields of a thermonuclear reactor are connected with a plurality of support beams. In a case where seismic vibrations are exerted to such a thermonuclear reactor, since the ends on the cutting side are coupled with the support beams, vibrations of the upper outer exhaust duct shields are greatly restricted. Thus, since there is no more such a possibility, for example, that an exhaust duct connected to the upper portion of a reactor main body is greatly distorted due to the seismic response of the upper outside exhaust duct shields to result in the failure of the connection portion with a vacuum pump, the release of tritium due to failure of the evacuating device can be prevented. (Yoshino, Y.)

  14. (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

  15. 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.)

  16. 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

  17. 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

  18. Solute diffusion in Pu and Ce

    International Nuclear Information System (INIS)

    Marbach, Gabriel; Charissoux, Christian; Janot, Christian

    1976-01-01

    The diffusion rate of Co, Au and Ag in the bcc phases of Pu and Ce is studied. In the bcc delta phase of Pu, Au and Ag have the same diffusion rate that the matrix and Co is a very rapid solute. In the bcc phase of Ce, the diffusion coefficients of the three metals are remarkably high. This phenomenon is also observed in certain metallic matrices (Pb, Sn, [fr

  19. Metamagnetism in Ce(Ga,Al)2

    Indian Academy of Sciences (India)

    Abstract. Effect of Al substitution on the magnetic properties of Ce(Ga1−x Alx )2 (x = 0, 0.1 and. 0.5) system has been studied. The magnetic state of CeGa2 is found to be FM with a TC of 8 K, whereas the compounds with x =0.1 and 0.5 are AFM and possess TN of about 9 K. These two com- pounds undergo metamagnetic ...

  20. 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

  1. Total catalytic wet oxidation of phenol and its chlorinated derivates with MnO2/CeO2 catalyst in a slurry

    OpenAIRE

    Luna, A. J.; Rojas, L. O. A.; Melo, D. M. A.; Benachour, M.; Sousa, J. F. de

    2009-01-01

    In the present work, a synthetic effluent of phenol was treated by means of a total oxidation process-Catalyzed Wet Oxidation (CWO). A mixed oxide of Mn-Ce (7:3), the catalyst, was synthesized by co-precipitation from an aqueous solution of MnCl2 and CeCl3 in a basic medium. The mixed oxide, MnO2/CeO2, was characterized and used in the oxidation of phenol in a slurry reactor in the temperature range of 80-130ºC and pressure of 2.04-4.76 MPa. A phenol solution containing 2.4-dichlorophenol and...

  2. Ce O2-Zr O2 tetragonal ceramics (Ce-TZP): mechanical properties

    International Nuclear Information System (INIS)

    Nono, Maria do Carmo de Andrade

    1994-01-01

    This work presents the development and the characterization of Ce O 2 -stabilized tetragonal Zr O 2 polycrystals (Ce-TZP ceramics), since it is considered candidate material for applications as structural high performance ceramics, an as substitute of some metallic materials. Ce-TZP ceramics attain remarkable increasing in strength and fracture toughness. Sintered ceramics were fabricated from mixtures of powders containing different Ce O 2 content prepared by conventional mechanical technique. It were adopted the bending strength, Vickers hardness and fracture toughness techniques to the determination of the mechanical parameters. These results were discussed and compared to those published in international literature. (author)

  3. 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)

  4. 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

  5. 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

  6. Neutronic reactor

    International Nuclear Information System (INIS)

    Carleton, J.T.

    1977-01-01

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment. 3 claims, 6 figures

  7. 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

  8. 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.

  9. 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

  10. 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.)

  11. 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.)

  12. Tetravalent Ce in the Nitrate-Decorated Hexanuclear Cluster [Ce 63 -O) 43 -OH) 4 ] 12+ : A Structural End Point for Ceria Nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Estes, Shanna L.; Antonio, Mark R.; Soderholm, L.

    2016-03-17

    We describe the synthesis and characterization of three glycine-stabilized hexanuclear Cely cluster compounds, each containing the [Ce-6(mu(3)-O)(4)(mu(3)-OH)(4)](12+) core structure. Crystallized from aqueous nitrate solutions with pH < 0, the core cluster structures exhibit variable decoration by nitrate, glycine, and water ligands depending on solution conditions, where increased nitrate and glycine decoration of the cluster core was observed for crystals synthesized at high Ce and nitrate concentrations. No other crystalline products were observed using this synthetic route. In addition to confirming the tetravalent oxidation state of cerium in one of the reported clusters, cyclic voltammetry also indicates that Ce-IV is reduced at similar to+0.60 V vs Ag/AgCl (3 M NaCl), which is significantly less than the standard electrode potential. This large decrease in the Ce-IV/Ce-III reduction potential suggests that Ce-IV is significantly stabilized relative to Ce-III within the examined cluster. These compounds are discussed in terms of their importance as small, end member, ceric oxide nanoparticles. Single-crystal structural solutions, together with voltammetry and electrolysis data, permit the decoupling of Ce-III defects and substoichiometry. In addition, Ce-Ce distances can be used to determine an "effective" CeO2-x lattice constant, providing a simple method for comparing literature descriptions. The results are discussed in terms of their potential implications for the mechanisms by which nanoparticle ceria serve as catalysts and oxygen-storage materials.

  13. Fate and effects of CeO2 nanoparticles in aquatic ecotoxicity tests.

    Science.gov (United States)

    Van Hoecke, Karen; Quik, Joris T K; Mankiewicz-Boczek, Joanna; De Schamphelaere, Karel A C; Elsaesser, Andreas; Van der Meeren, Paul; Barnes, Clifford; McKerr, George; Howard, C Vyvyan; Van de Meent, Dik; Rydzyński, Konrad; Dawson, Kenneth A; Salvati, Anna; Lesniak, Anna; Lynch, Iseult; Silversmit, Geert; De Samber, Björn; Vincze, Laszlo; Janssen, Colin R

    2009-06-15

    Cerium dioxide nanoparticles (CeO2 NPs) are increasingly being used as a catalyst in the automotive industry. Consequently, increasing amounts of CeO2 NPs are expected to enter the environment where their fate in and potential impacts are unknown. In this paper we describe the fate and effects of CeO2 NPs of three different sizes (14, 20, and 29 nm) in aquatic toxicity tests. In each standard test medium (pH 7.4) the CeO2 nanoparticles aggregated (mean aggregate size approximately 400 nm). Four test organisms covering three different trophic levels were investigated, i.e., the unicellular green alga Pseudokirchneriella subcapitata, two crustaceans: Daphnia magna and Thamnocephalus platyurus, and embryos of Danio rerio. No acute toxicity was observed for the two crustaceans and D. rerio embryos, up to test concentrations of 1000, 5000, and 200 mg/L, respectively. In contrast, significant chronic toxicity to P. subcapitata with 10% effect concentrations (EC10s) between 2.6 and 5.4 mg/L was observed. Food shortage resulted in chronic toxicity to D. magna, for wich EC10s of > or = 8.8 and or = 0.052 and effect of dissolved Ce or CeO2 NP uptake or adsorption, nor to an indirect effect of nutrient depletion (by sorption to NPs) or physical light restriction (through shading by the NPs). However, observed clustering of NPs around algal cells may locally cause a direct or indirect effect.

  14. 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.

  15. 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.)

  16. 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

  17. 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

  18. 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

  19. Reactor container

    International Nuclear Information System (INIS)

    Otsuka, Hiroaki; Yoshida, Takashi.

    1979-01-01

    Purpose: To prevent rain water falling along the outer wall of the container during the construction work of an atomic power plant from making ingress into the inner part of a reactor container through a large size material carry-in port. Constitution: A weir for preventing the ingress of rain water is provided on the border between the foot floor of a large material carry-in port provided on the side surface at the bottom part of the reactor container and the floor surface of the building. This weir is of a semi-circular plate shape, and formed so that the lower semi-circular part of the carry-in port is tightly closed. (Kamimura, M.)

  20. 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.)

  1. 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.)

  2. 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.

  3. 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

  4. 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.)

  5. 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.)

  6. 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.)

  7. Cyclic voltammetry study of Ce(IV/Ce(III redox couple and Ce(IV-F complex in sulfuric acid medium

    Directory of Open Access Journals (Sweden)

    J. G. He

    2016-10-01

    Full Text Available In this paper the electrochemical behaviors of Ce(IV/Ce(III redox couple and Ce(IV - F complex in sulfuric acid medium were studied by cyclic voltammetry using a platinum electrode. Both of the Ce(IV/Ce(III couple in Ce(IV solution and Ce(IV - F complex is a quasi-reversible process, and gives a linear correlation between the peak potentials and square root of scan rates, showing that the kinetics of the overall process is diffusion controlled. The complexation of cerium(IV and fluoride is favorable for the oxidation of Ce(III. The kinetic parameters such as diffusion coefficients, anodic transfer coefficients and rate constants were studied.

  8. 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.

  9. Renormalized electronic structures of CeSi{sub 2}, CeRu{sub 2} and CeAl{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Costa-Quintana, J. [Universitat Autonoma de Barcelona (Spain). Grup d`Electromagnetisme; Gonzalez-Leon, E. [Universitat Autonoma de Barcelona (Spain). Grup d`Electromagnetisme; Lopez-Aguilar, F. [Universitat Autonoma de Barcelona (Spain). Grup d`Electromagnetisme; Puig-Puig, L. [Universitat Autonoma de Barcelona (Spain). Grup d`Electromagnetisme; Sanchez-Lopez, M.M. [Universitat Autonoma de Barcelona (Spain). Grup d`Electromagnetisme

    1995-02-01

    The renormalized density of states of some Ce compounds is analyzed by considering self-energy effects. We study the influence of the hybridization introduced by the self-energy and how it can affect the shape of the characteristic lower, upper and middle-energy resonance. ((orig.)).

  10. 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

  11. 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

  12. 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

  13. 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

  14. Ce-valence state and hydrogen-induced volume effects in Ce-based intermetallic compounds and their hydrides

    International Nuclear Information System (INIS)

    Stange, M.; Paul-Boncour, V.; Latroche, M.; Percheron-Guegan, A.; Isnard, O.; Yartys, V.A.

    2005-01-01

    An average Ce-valence state (v) of two types of Ce-containing intermetallic compounds, equiatomic CeNiX (X= Al, Ga, Sn) and CeM 3 (M= Ni, Co, Mn), and their hydrides was estimated from X-ray absorption spectroscopy (XAS) and analysed in parallel with hydrogen-induced volume changes on hydrogenation. The largest valence states in the initial compounds were found in CeM 3(v=3.32-3.36) followed by CeNiAl 1-x Ga x (v=3.25-3.27). This contrasts to CeNiSn, which is close to a pure trivalent state (v=3.07). On hydrogenation, a conversion from mixed-valent CeNiAl 1-x Ga x , x=0.5, 1, to pure Ce III hydrides takes place (ΔV /V=19.6-20.0 %). In CeNiSnD z (z=1, 1.8, ΔV/V=3.0, 8.0%) the changes in the valence state towards Ce III are very small. The situation for the CeM 3 -hydrides is complex. For CeNi 3 D 2.8 , CeNi 2.75 Mn 0.25 D 3.4 and CeCo 3 D 3.4 (ΔV /V=24-32%) where rather similar electronic properties can be expected, a decrease in the contribution of Ce IV for CeNi 3 D 3 and CeNi 2.75 Mn 0.25 D 3.4 (v=3.18 and 3.12, respectively) contrasts to the behaviour of CeCo 3 D 3.4 where the hydrogen induced valence change is very small (v=3.32)

  15. Photocatalytic reactor

    Science.gov (United States)

    Bischoff, Brian L.; Fain, Douglas E.; Stockdale, John A. D.

    1999-01-01

    A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

  16. 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

  17. 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)

  18. 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)

  19. 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.

  20. 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

  1. Metamagnetism in Ce(Ga,Al)2

    Indian Academy of Sciences (India)

    Metamagnetism in Ce(Ga,Al)2. K G SURESH1,∗. , S RADHA2 and A K NIGAM3. 1Department of Physics, Indian Institute of Technology, Mumbai 400 076, India. 2Department of Physics, Mithibai College, Mumbai 400 056, India. 3Tata Institute of Fundamental Research, Homi Bhabha Road, Mumbai 400 005, India. ∗.

  2. (Ln = Ce, Pr and Nd) microwave ceramics

    Indian Academy of Sciences (India)

    125–128. c Indian Academy of Sciences. Influence of lead oxide addition on LnTiTaO6 (Ln = Ce, Pr and Nd) ... A number of samples with improved microwave dielectric properties were obtained on all the systems suitable for practical ... circuits due to their compactness, thermal stability, low cost of production, high efficiency ...

  3. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  4. 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.)

  5. 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.

  6. 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.)

  7. Reactor container

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-07

    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.).

  8. 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.)

  9. 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.

  10. Crystal structures of isotypic aluminides CeRuAl and CeRhAl

    International Nuclear Information System (INIS)

    Gribanov, A.V.; Tursina, A.I.; Grytsiv, A.V.; Murashova, E.V.; Bukhan'ko, N.G.; Rogl, P.; Seropegin, Y.D.; Giester, G.

    2008-01-01

    The crystal structures of the compounds CeRuAl and CeRhAl were studied by single crystal X-ray diffraction. Both compounds crystallize in the orthorhombic LaNiAl type with space group Pnma (No. 62) and are related to the structure types of hexagonal ZrNiAl and orthorhombic TiNiSi. Parameters for CeRuAl and CeRhAl are: a = 7.2057(2) A, b = 4.0589(1) A, c = 15.8728(5) A; R F = 0.023; a = 7.107(2) A, b = 4.2237(10) A, c = 15.965(3) A; R F 0.032, respectively

  11. OILY WASTEWATER TREATMENT FOR CATALYTIC OXIDATION WITH K-MNO2/CeO2

    OpenAIRE

    JERÔNIMO, Carlos Enrique de Medeiros; FERNANDES, Hermano Gomes; SOUSA, João Fernandes

    2012-01-01

    The search for alternatives for the treatment of industrial effluents, especially high toxicity is one of the main difficulties of the industrial treatment systems. This work aims to use potassium as a promoter of electrons for the oxidation to the catalyst by impregnating it MnO2/CeO2 to degrade the effluent coming from the processing of cashew nuts in the presence of oxygen. The experiments were performed in a slurry bed reactor. In the experiments were kept constant initial concentratio...

  12. 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.

  13. R-ES-ONAN--CE--IO

    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 ...

  14. 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)

  15. 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

  16. NEUTRONIC REACTOR

    Science.gov (United States)

    McGarry, R.J.

    1958-04-22

    Fluid-cooled nuclear reactors of the type that utilize finned uranium fuel elements disposed in coolant channels in a moderater are described. The coolant channels are provided with removable bushings composed of a non- fissionable material. The interior walls of the bushings have a plurality of spaced, longtudinal ribs separated by grooves which receive the fins on the fuel elements. The lands between the grooves are spaced from the fuel elements to form flow passages, and the size of the now passages progressively decreases as the dlstance from the center of the core increases for the purpose of producing a greater cooling effect at the center to maintain a uniform temperature throughout the core.

  17. 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

  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. SLiCE: a novel bacterial cell extract-based DNA cloning method

    Science.gov (United States)

    Zhang, Yongwei; Werling, Uwe; Edelmann, Winfried

    2012-01-01

    We describe a novel cloning method termed SLiCE (Seamless Ligation Cloning Extract) that utilizes easy to generate bacterial cell extracts to assemble multiple DNA fragments into recombinant DNA molecules in a single in vitro recombination reaction. SLiCE overcomes the sequence limitations of traditional cloning methods, facilitates seamless cloning by recombining short end homologies (≥15 bp) with or without flanking heterologous sequences and provides an effective strategy for directional subcloning of DNA fragments from Bacteria Artificial Chromosomes (BACs) or other sources. SLiCE is highly cost effective as a number of standard laboratory bacterial strains can serve as sources for SLiCE extract. In addition, the cloning efficiencies and capabilities of these strains can be greatly improved by simple genetic modifications. As an example, we modified the DH10B Escherichia coli strain to express an optimized λ prophage Red recombination system. This strain, termed PPY, facilitates SLiCE with very high efficiencies and demonstrates the versatility of the method. PMID:22241772

  20. SLiCE: a novel bacterial cell extract-based DNA cloning method.

    Science.gov (United States)

    Zhang, Yongwei; Werling, Uwe; Edelmann, Winfried

    2012-04-01

    We describe a novel cloning method termed SLiCE (Seamless Ligation Cloning Extract) that utilizes easy to generate bacterial cell extracts to assemble multiple DNA fragments into recombinant DNA molecules in a single in vitro recombination reaction. SLiCE overcomes the sequence limitations of traditional cloning methods, facilitates seamless cloning by recombining short end homologies (≥15 bp) with or without flanking heterologous sequences and provides an effective strategy for directional subcloning of DNA fragments from Bacteria Artificial Chromosomes (BACs) or other sources. SLiCE is highly cost effective as a number of standard laboratory bacterial strains can serve as sources for SLiCE extract. In addition, the cloning efficiencies and capabilities of these strains can be greatly improved by simple genetic modifications. As an example, we modified the DH10B Escherichia coli strain to express an optimized λ prophage Red recombination system. This strain, termed PPY, facilitates SLiCE with very high efficiencies and demonstrates the versatility of the method.

  1. 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

  2. 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

  3. Neutron scattering from α-Ce at epithermal neutron energies

    Indian Academy of Sciences (India)

    tures (~10 K), whereas CeSn3 is an α-Ce-like (valence fluctuation) system (TK ~ 35. meV) that remains paramagnetic down to low temperatures. Figure 1 shows the magnetic response from CeIn3 at 20 K [9]. A broad, well-defined crystal field exci- tation and a central quasielastic distribution are observed in this compound.

  4. Facile hydrothermal synthesis of CeO2 nanopebbles

    Indian Academy of Sciences (India)

    Administrator

    Abstract. 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.

  5. 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)

  6. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    Sklenka, L.; Kropik, M.

    2006-01-01

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

  7. Metamagnetism in Ce (Ga, Al) 2

    Indian Academy of Sciences (India)

    Strongly Correlated Electron Systems Volume 58 Issue 5-6 May-June 2002 pp 769-771 ... The magnetic state of CeGa2 is found to be FM with a C of 8 K, whereas the compounds with =0.1 and 0.5 are AFM and possess N of about 9 K. These two compounds undergo metamagnetic transition and the critical fields are ...

  8. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

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

  9. Survey of research reactors

    International Nuclear Information System (INIS)

    Boek, H.; Villa, M.

    2004-06-01

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

  10. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  11. Two-orbital model for CeB6

    Science.gov (United States)

    Singh, Dheeraj Kumar

    2017-05-01

    We describe a two-orbital tight-binding model with bases belonging to the Γ8 quartet. It captures several characteristics of the Fermiology unravelled by the recent angle-resolved photoemission spectroscopic (ARPES) measurements on cerium hexaboride CeB6 samples cleaved along different high-symmetry crystallographic directions, which includes the ellipsoid-like Fermi surfaces (FSs) with major axes directed along Γ-X. We calculate various multipolar susceptibilities within the model and identify the susceptibility that shows the strongest divergence in the presence of standard onsite Coulomb interactions and discuss its possible implication and relevance with regard to the signature of strong ferromagnetic correlations existent in various phases as shown by the recent experiments.

  12. Economic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Owen, P.S.; Parker, M.B.; Omberg, R.P.

    1979-05-01

    The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U 3 O 8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented

  13. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-01-01

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

  14. 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.

  15. 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

  16. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  17. Determining Reactor Neutrino Flux

    OpenAIRE

    Cao, Jun

    2011-01-01

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understa...

  18. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  19. A Concept Design of Reactor Protection System for PALLAS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Young Rok; Kim, Sung Hun; Kim, Jung Seon [Control Tech. Research Inst., Ulsan (Korea, Republic of); Park, Geun Ok; Koo, In Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Hee Seok; Cho, Gi Ho [OPENCULTURE Co., Daejeon (Korea, Republic of)

    2009-05-15

    The Reactor Protection System (RPS) is classified as a Safety Category 1 (SC1) I and C systems and complies with IEEE standards for Class 1E equipment. The RPS consists of First Reactor Protection System (FRPS) and Second Reactor Protection System (SRPS). There are two independent and diverse RPS to safely shutdown the reactor under all circumstances. The FRPS generates signals to actuate reactor trip by using CEA Bank1 automatically whenever monitored process variables reach predefined limits. The SRPS generates signals to actuate reactor trip by using Second Shutdown System (SSS) absorber assemblies automatically whenever monitored process variables reach predefined limits. The SRPS is an analogue system while the FRPS is a digital system.

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  1. Reactor container

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo; Yanai, Ryoichi.

    1976-01-01

    Object: To provide a reactor container which is free from water shock action or condensing vibrations and cannot be readily broken by a missile from a pump impeller, pipe whipping, steam jet reaction, etc., and which also quickly condenses issuing steam and possesses a large vibration-proof strength. Structure: A high pressure containment vessel accommodating a pressure container includes a plurality of pressurized water tanks arranged along its inner periphery, and a pneumatic valve is provided in a lower portion of each of these pressurized water tanks. If an accident occurs, vapor is caused to issue from the pressure container into the vessel. When a certain value is reached, the pneumatic valves are opened, whereby the gas within the pressurized water tanks causes pressurized water to flow through the pipe and be ejected from spray nozzles to cause condensation of water within the vessel. Further, water of a pool within the container is circulated to allow heat release to the outside. (Horiuchi, T.)

  2. Research reactor safety - an overview of crucial aspects

    International Nuclear Information System (INIS)

    Laverie, M.

    1998-01-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  3. Spin reorientation and Ce-Mn coupling in antiferromagnetic oxypnictide CeMnAsO

    Science.gov (United States)

    Vaknin, David; Zhang, Qiang; Peterson, Spencer; Dennis, Kevin; Tian, Wei

    2015-03-01

    Structure and complex magnetic properties of CeMnAsO, a parent compound of the ``1111''-type oxypnictides, have been investigated using neutron powder diffraction and magnetization measurements. Whereas there is no structural transition from the P4/nmm tetragonal phase below 420 K, CeMnAsO undergoes a C-type antiferromagnetic order with Mn2+ (S = 5 / 2) moments pointing along the c-axis below a relatively high Néel temperature of TN = 345 K. Below TSR = 35 K, two instantaneous transitions occur where the Mn moments reorient to the ab-plane preserving the C-type magnetic order, and Ce moments undergo long-range AFM ordering with moments in the ab-plane. Another transition to a noncollinear magnetic structure occurs below 7 K. We find that CeMnAsO primarily falls into the category of a local-moment antiferromagnetic insulator in which the nearest-neighbor interaction (J1) is dominant. The spin reorientation transition driven by the coupling between rare earth Ce and transition metal seems to be common to Mn, Fe and Cr ion,but not to Co and Ni ions in the iso-structural oxypnictides. Supported by the Office of Basic Energy Sciences, US-DOE, Number DE-AC02-07CH11358.

  4. Structure of Nano-sized CeO2 Materials: Combined Scattering and Spectroscopic Investigations.

    Science.gov (United States)

    Marchbank, Huw R; Clark, Adam H; Hyde, Timothy I; Playford, Helen Y; Tucker, Matthew G; Thompsett, David; Fisher, Janet M; Chapman, Karena W; Beyer, Kevin A; Monte, Manuel; Longo, Alessandro; Sankar, Gopinathan

    2016-11-04

    The structure of several nano-sized ceria, CeO 2 , systems was investigated using neutron and X-ray diffraction and X-ray absorption spectroscopy. Whilst both diffraction and total pair distribution functions (PDFs) revealed that in all of the samples the occupancy of both Ce 4+ and O 2- are very close to the ideal stoichiometry, the analysis using Reverse Monte Carlo technique revealed significant disorder around oxygen atoms in the nano-sized ceria samples in comparison to the highly crystalline NIST standard. In addition, the analysis revealed that the main differences observed in the pair correlations from various X-ray and neutron diffraction techniques were attributable to the particle size of the CeO 2 prepared by the reported three methods. Furthermore, detailed analysis of the Ce L 3 - and K-edge EXAFS data support this finding; in particular the decrease in higher shell coordination numbers with respect to the NIST standard, is attributed to differences in particle size. © 2016 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  5. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

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

  6. Developments in the regulation of research reactors

    International Nuclear Information System (INIS)

    Loy, J.

    2003-01-01

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

  7. [Effects of Arbuscular Mycorrhizal Fungi on the Growth and Ce Uptake of Maize Grown in Ce-contaminated Soils].

    Science.gov (United States)

    Wang, Fang; Guo, Weil; Ma, Peng-kun; Pan, Liang; Zhang, Jun

    2016-01-15

    A greenhouse pot experiment was conducted to investigate the effects of arbuscular mycorrhizal (AM) fungi Glomus aggregatum (GA) and Funneliformis mosseae (FM) on AM colonization rate, biomass, nutrient uptake, C: N: P stoichiometric and Ce uptake and transport by maize (Zea mays L.) grown in soils with different levels of Ce-contaminated (100, 500 and 1000 mg x kg(-1)). The aim was to provide basic data and technical support for the treatment of soils contaminated by rare earth elements. The results indicated that symbiotic associations were successfully established between the two isolates and maize, and the average AM colonization rate ranged from 7. 12% to 74.47%. The increasing concentration of Ce in soils significantly decreased the mycorrhizal colonization rate, biomass, nutrition contents and transport rate of Ce from root to shoot of maize, and significantly increased C: P and N: P ratios and Ce contents in shoot and root of maize. Both AM fungi inoculations promoted the growth of maize, but the promoting role of FM was more significant than that of GA in severe Ce-contaminated soils. There were no significant differences in the growth of maize between two AM fungi in mild and moderate Ce-contaminated soils. Inoculation with AM fungi significantly improved nutritional status of maize by increasing nutrient uptake and decreasing C: N: P ratios. GA was more efficient than FM in enhancing nutrient uptake in mild and moderate Ce-contaminated soils, while FM was more efficient in severe Ce-contaminated soils. Moreover, inoculation with AM fungi significantly increased Ce contents of shoot and root in mild Ce-contaminated soils, but had no significant effect on Ce contents of maize in moderate and severe Ce-contaminated soils, and promoted the transport of Ce from root to shoot. The experiment demonstrates that AM fungi can alleviate toxic effects of Ce on plants and have a potential role in the phytoremediation of soils contaminated by rare earth elements.

  8. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

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

  9. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    Acosta Ezcurra, T.; Garcia Rodriguez, B.M.

    1996-01-01

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

  10. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  11. Spin reorientation and Ce-Mn coupling in antiferromagnetic oxypnictide CeMnAsO

    OpenAIRE

    Zhang, Qiang; Tian, Wei; Peterson, Spencer G.; Dennis, Kevin W.; Vaknin, David

    2014-01-01

    Structure and magnetic properties of high-quality polycrystlline CeMnAsO, a parent compound of the "1111"-type oxypnictides, have been investigated using neutron powder diffraction and magnetization measurements. We find that CeMnAsO undergoes a C-type antiferromagnetic order with Mn$^{2+}$ ($S=5/2$) moments pointing along the \\textit{c}-axis below a relatively high N\\'{e}el temperature of $T_{\\rm N} = 345$ K. Below $T_{\\rm SR} = 35$ K, two instantaneous transitions occur where the Mn moments...

  12. Neutron emission following muon capture in Ce-142, Ce-140, Ba-138, and Sn-120.

    Science.gov (United States)

    Lucas, G. R., Jr.; Martin, P.; Welsh, R. E.; Jenkins, D. A.; Powers, R. J.; Kunselman, A. R.; Miller, G. H.

    1973-01-01

    Branching ratios to excited nuclear states formed after muon capture have been measured with Ge(Li) detectors. The delayed gamma rays were observed in studies of muonic Ce-142, Ce-140, Ba-138, and Sn-120, using separated isotopes. The resulting isotopes formed indicate at least a 60% probability of neutron emission upon muon capture, with the most likely product resulting from single-neutron emission. No evidence for delayed proton emission with a probability higher than 2% was found. Using our more precise energies for the observed nuclear transitions, we present revised energy levels schemes for La-141, La-139, Cs-137, and In-119.

  13. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  14. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  15. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  16. HTGR type reactors for the heat market

    International Nuclear Information System (INIS)

    Oesterwind, D.

    1981-01-01

    Information about the standard of development of the HTGR type reactor are followed by an assessment of its utilization on the heat market. The utilization of HTGR type reactors is considered suitable for the production of synthesis gas, district heat, and industrial process heat. A comparison with a pit coal power station shows the economy of the HTGR. Finally, some aspects of introducing new technologies into the market, i.e. small plants in particular are investigated. (UA) [de

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

    International Nuclear Information System (INIS)

    Amusai, J.H.; Schandorf, C.; Yeboah, J.

    2001-01-01

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

  18. 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

  19. Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Nohara, Kiyohiko

    2009-01-01

    The structural material is one of key issues for the development of reliable superconducting magnets and peripheral equipments of fusion reactors. Standard stainless steels like SUS 304 and 316 steels available at present do not meet requirements. We are developing a new austenitic steel that has proposed target properties named 'JAERI BOX'. Additions of N and V at different amounts were tested to improve strength and fracture toughness of a base alloy SUS316LN at 4.2 K. Mechanical properties of the developed steel were examined. It is found that the charpy absorbed energy and the fracture toughness of the developed steel at 4.2 K are within JAERI BOX. (T.I.)

  20. Synthesis and characterization of Ce doped MFI zeolite

    Energy Technology Data Exchange (ETDEWEB)

    Kalita, Banani [Department of Chemistry, Gauhati University, Guwahati, Assam 781014 (India); Talukdar, Anup K., E-mail: anup_t@sify.com [Department of Chemistry, Gauhati University, Guwahati, Assam 781014 (India)

    2012-04-16

    Highlights: Black-Right-Pointing-Pointer Cerium was incorporated into the tetrahedral position of MFI zeolite structure. Black-Right-Pointing-Pointer Unit cell volume increases with an increase of Ce content in the framework of MFI. Black-Right-Pointing-Pointer A band at 310 nm in the UV-vis spectra indicates Ce incorporation in MFI structure. Black-Right-Pointing-Pointer 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.

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, B.

    1998-10-19

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

  3. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

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

  4. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

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

  5. Reactor calculations for improving utilization of TRIGA reactor

    International Nuclear Information System (INIS)

    Ravnik, M.

    1986-01-01

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

  6. 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)

  7. Dissolution of Ce from Cd Solution Containing U/Ce Elements by Electrolysis

    International Nuclear Information System (INIS)

    Kim, Si Hyung; Kim, Gha-Young; Lee, Seung-jai; Kim, Taek-Jin; Paek, Seungwoo; Ahn, Do-Hee

    2015-01-01

    The U-TRU metal alloy can be supplied by the Pyroprocessing, specifically UTRU recovery process using liquid cadmium cathode (LCC). In a certain case, a lot of rare earth (RE) element could be recovered on the LCC with the TRU element during the Pyroprocessing when the concentration of RE ions is higher than that of the TRU ions in the salt. In this case, most of the RE element needs to be removed from the Cd solution containing U/TRU/RE elements. RAR(Residual Actinides Recovery) technique used the mixed electrolytic-chemical process. In this study, only electrolysis technique was utilized to remove Ce element from Cd solution containing U/Ce elements. U-TRU alloy having less impurity is necessary for the fabrication of SFR fuel and these U-TRU elements can be prepared by Pyroprocessing. Electrolytic method was used to reduce the amount of Ce elements from the Cd solution containing U/Ce elements. It is judged from this study that electrolytic dissolution can be one of the methods to reduce RE elements from the Cd solution containing U-TRU-RE elements

  8. Applications of an Automated and Quantitative CE-Based Size and Charge Western Blot for Therapeutic Proteins and Vaccines.

    Science.gov (United States)

    Rustandi, Richard R; Hamm, Melissa; Lancaster, Catherine; Loughney, John W

    2016-01-01

    Capillary Electrophoresis (CE) is a versatile and indispensable analytical tool that can be applied to characterize proteins. In recent years, labor-intensive SDS-PAGE and IEF slab gels have been replaced with CE-SDS (CGE) and CE-IEF methods, respectively, in the biopharmaceutical industry. These two CE-based methods are now an industry standard and are an expectation of the regulatory agencies for biologics characterization. Another important and traditional slab gel technique is the western blot, which detects proteins using immuno-specific reagents after SDS-PAGE separation. This technique is widely used across industrial and academic laboratories, but it is very laborious, manual, time-consuming, and only semi-quantitative. Here, we describe the applications of a relatively new CE-based western blot technology which is automated, fast, and quantitative. We have used this technology for both charge- and size-based CE westerns to analyze biotherapeutic and vaccine products. The size-based capillary western can be used for fast antibody screening, clone selection, product titer, identity, and degradation while the charge-based capillary western can be used to study product charge heterogeneity. Examples using this technology for monoclonal antibody (mAb), Enbrel, CRM197, and Clostridium difficile (C. difficile) vaccine proteins are presented here to demonstrate the utility of the capillary western techniques. Details of sample preparation and experimental conditions for each capillary western mode are described in this chapter.

  9. Public Data at CeSAM

    Science.gov (United States)

    Moreau, C.; Le Brun, V.; Agneray, F.; Gimenez, S.; Roehlly, Y.

    2014-05-01

    Modern large observational programs produce important amounts of data from various origins, and need high level data-quality control, fast data access via easy-to-use graphic interfaces, as well as possibility to cross-correlate informations coming from different observations. The Centre de donneS Astrophysique de Marseille (CeSAM) has for mission to provide support to the teams in charge of the observational programs (specifications, software development and infrastructures), to produce and give the final data by proposing tools increasing their scientific value. We present here the various datasets hosted by the CeSAM under the ANIS environment. Even if mainly oriented towards large spectro-photometric extragalactic samples, we host as well an exoplanet transit database, or resolved galaxy data. The main datasets are those from the VVDS final release (45000 spectra), the Herschel extragalactic surveys (HerMES, HRS, GOODS-Herschel, VNGS), the CFHTLS-WIDE photometric redshifts catalog (17 millions objects), or the HST-COSMOS information system, that allows request from any of the 6 and access to the corresponding values in the other catalogs.

  10. PNAE and the Mandala System in the municipality of MAURITI/CE (Brazil)

    OpenAIRE

    Tiago Cartaxo de Lucena; Nataniele dos Santos Alencar; Jamily Freire Gonçalves; João Cesar Abreu de Oliveira; José Levi Furtado Sampaio

    2017-01-01

    The National School Feeding Program (PNAE) in Mauriti/CE (Brazil) has a positive impact on family farming, as it encourages and promotes the commercialization of its products. The goal of the program is to benefit students in basic education with healthy and nutritious food. It is worth mentioning that the PNAE is interconnected to the mandala system, whose principle is environmental sustainability and the improvement of the food standard, consequently improving the income and quality of life...

  11. Development of Seismic Resistance Position Indicator for the Integral Reactor

    International Nuclear Information System (INIS)

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

    2008-01-01

    The present paper is related to position sensing means and more particularly, to a magnetic position sensor using a permanent magnet and a compact arrangement of reed switches in a nuclear power plant. 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. Its indicating resolution (1-1/2 inch (38.1mm)) is suitable to measure the position of a control rod which is driven by a motor having steps of 3/4 inch (19.05mm). But 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. These design features require a CEDM for the integral reactor to have a fine-step movement for a fine reactivity control. 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. The control rod position indicator having a seismic resistance characteristic for the integral reactor was developed on the basis of the RSPT technology identified through the survey

  12. Research reactor instrumentation

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2001-02-01

    This is a textbook on research reactor instrumentation for training purposes, it gives a survey on research reactor instrumentation requirements and eight exercises covering the major aspects of this topic are presented. (author)

  13. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  14. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

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

  15. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  16. Reactor shutdown method

    International Nuclear Information System (INIS)

    Nishino, Yoshitaka; Sawa, Toshio; Matsumoto, Takayuki; Osumi, Katsumi; Usui, Naoshi.

    1991-01-01

    A device for injecting a hydrogen gas, a chelating agent or a reducing agent is disposed in a reactor water recycling system. Upon reactor shutdown, the hydrogen gas, the chelating agent or the reducing agent is injected to primary coolants. With such a procedure, radioactive ions formed by the dissolution of oxide layers at the surface of pipelines and equipments in a reactor water recycling system and a reactor water cleanup system are removed from the primary coolants by a reactor water cleanup device. Accordingly, since the dose rate at the surface of the pipelines can be reduced, the operator's radiation dose can be reduced upon periodical inspection for a power plant. Further, the inner pressure of the reactor is kept higher than the saturated steam pressure at the reactor water temperature to suppress boiling of the reactor water. This can suppress the peeling of cruds deposited to the surface of the fuel cladding tube. (I.N.)

  17. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

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

  18. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

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

  19. 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.

  20. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  1. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  2. Neutron scattering from α-Ce at epithermal neutron energies

    Indian Academy of Sciences (India)

    the constant-ω integration within an energy bandwidth of ±30 meV around. 170 meV [13]. The continuous curve through the data shows the Ce3+ mag- netic form factor. The doted line shows the calculated form factor for an itinerant 4f electron band for α-Ce [14]. et al [14] calculated within the itinerant 4f band model for Ce.

  3. On Line Neutron Flux Mapping in Fuel Coolant Channels of a Research Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Villard, Jean-Francois; Destouches, Christophe; Braoudakis, George; Wassink, David; Sinclair, Bradley; Osborn, John-C.; Wu, Huayou; Blandin, C.; Thevenin, Mathieu; Corre, Gwenole; Normand, Stephane

    2013-06-01

    This work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E ∼0.97). (authors)

  4. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

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

  5. Ulysse, mentor reactor

    International Nuclear Information System (INIS)

    Bouquin, B.; Rio, I.; Safieh, J.

    1997-01-01

    On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

  6. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  7. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  8. The Jules Horowitz reactor

    International Nuclear Information System (INIS)

    2003-01-01

    The Jules Horowitz reactor is the future european reactor for irradiation. It will be used for materials and new fuels irradiation. Experiments for the safety and the validation of neutronics calculation will be also realized. This paper presents the design and the performance of the reactor and the schedule of the remaining design studies. (A.L.B.)

  9. 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

  10. High-Resolution Resonance Photoemission Study of CeNi

    International Nuclear Information System (INIS)

    Sekiyama, A.; Kadono, K.; Iwasaki, T.; Imada, S.; Kasai, S.; Suga, S.; Araki, S.; Onuki, Y.

    2003-01-01

    We have performed the high-resolution Ce 3d - 4f resonance photoemission study of a considerably hybridized CeNi with the Kondo temperature of ∼ 150 K. The tail of the Kondo-resonance peak is predominantly observed in the bulk Ce 4f photoemission spectra, where its spin-orbit partner is suppressed compared with the so far reported surface-sensitive Ce 4d-4f resonance spectrum. Our results show that the bulk 4f electronic states are essentially understood by the single impurity Anderson model. (author)

  11. 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.

  12. 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

  13. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

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

  14. PRISM: An innovative liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

    1986-01-01

    This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design

  15. Study of the crystal field in CeF3 and CeF3:Pr3+

    Directory of Open Access Journals (Sweden)

    Shafikov F.F., Savinkov A.V., Tagirov M.S.

    2016-12-01

    Full Text Available The crystal field analysis based on calculations in the framework of the semi phenomenological exchange charge model was carried out. The set of crystal field parameters for Ce3+ and Pr3+ ions in the matrix CeF3 related to the crystallographic system of coordinates has been obtained and used to reproduce satisfactory the crystal field energies of Ce3+ and Pr3+ ions.

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

    International Nuclear Information System (INIS)

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

    1978-05-01

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

  17. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    OpenAIRE

    Koroush Shirvan; Mujid Kazimi

    2016-01-01

    A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and ...

  18. Unique differences in applying safety analyses for a graphite moderated, channel reactor

    International Nuclear Information System (INIS)

    Moffitt, R.L.

    1993-06-01

    Unlike its predecessors, the N Reactor at the Hanford Site in Washington State was designed to produce electricity for civilian energy use as well as weapons-grade plutonium. This paper describes the major problems associated with applying safety analysis methodologies developed for commercial light water reactors (LWR) to a unique reactor like the N Reactor. The focus of the discussion is on non-applicable LWR safety standards and computer modeling/analytical variances of standards. The approaches used to resolve these problems to develop safety standards and limits for the N Reactor are described

  19. Thorium fueled reactor

    Science.gov (United States)

    Sipaun, S.

    2017-01-01

    Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

  20. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  1. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

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

  2. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    Akatsu, Eiko

    1981-12-01

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

  3. 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.

  4. 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

  5. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  6. Architecture dependent availability analysis of RPS for Research Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Khalilur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Youngki; Park, Jaekwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250- 2.0 MW power rating or 2.5-10 Χ 10{sup 11} η/cm{sup 2}. s flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 10{sup 13} η/cm{sup 2}. s are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor. The aim of this research, in this article, was to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and their fault trees were developed using AIMS PSA software to get the unavailability. This is a starting point of attempt towards the standardization of I and C architecture for low and medium power research reactors.

  7. Spinning fluids reactor

    Science.gov (United States)

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

    2012-11-20

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

  8. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  9. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  10. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  11. Visual examination in nuclear reactor inservice inspection

    International Nuclear Information System (INIS)

    Kornvik, L.A.

    1985-01-01

    Visual examination is an important inspection method for nuclear reactors. New developments in TV technology give new possibilities for inspections that contribute to the economic and safe operation of nuclear power plants. As a supplier of nuclear boiling water reactors, ASEA-ATOM is constantly following-up its delivered plants by inspecting different parts of the reactor system after various periods of service. Inspections include both standard NDT-methods and different visual examination methods. TV inspection often offers advantages over other methods. Special developments such as the use of color TV, miniature TV cameras, development of stereo TV and the use of miniature remote controlled vehicles greatly enhance the usefulness and applicability of visual examination. This will for example, make it possible to make more adequate evaluation of indications and to take direct in-picture measurements. It will also give added possibilities to inspect reactor internals with respect to possible cracking or other defects

  12. The advanced test reactor strategic evaluation program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1989-01-01

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

  13. Cooperation in reactor design evaluation and licensing

    Energy Technology Data Exchange (ETDEWEB)

    Kaufer, B.; Wasylyk, A. [World Nuclear Association, London (United Kingdom)

    2014-07-01

    In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

  14. 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

  15. The role of 18F-FDG-PET/ceCT in peritoneal mesothelioma.

    Science.gov (United States)

    Dubreuil, Julien; Giammarile, Francesco; Rousset, Pascal; Rubello, Domenico; Bakrin, Naoual; Passot, Guillaume; Isaac, Sylvie; Glehen, Olivier; Skanjeti, Andrea

    2017-04-01

    The aim of this study was to assess glucose metabolism of multicystic peritoneal mesothelioma and epithelioid peritoneal mesothelioma by fluorine-18 fluorodeoxyglucose (F-FDG)-PET/contrast-enhanced computed tomography (ceCT) and to assess its prognostic impact. Twenty-three (14 women) patients, without previous treatment, underwent F-FDG-PET/ceCT before peritoneal mesothelioma cytoreductive surgery and intraperitoneal chemotherapy. F-FDG-PET/ceCT was interpreted prospectively as positive or negative. Maximum standardized uptake value (SUVmax) of each lesion was measured retrospectively on the basis of postsurgery data. At laparotomy, disease extension was estimated with the Peritoneal Cancer Index. The median follow-up was 27 months (95% confidence interval: 12.9-37.8); progression-free survival (PFS) was recorded. Nine patients were affected by multicystic and 14 were affected by epithelioid peritoneal mesothelioma. PET showed mild focal uptake in one case of multicystic peritoneal mesothelioma, whereas in eight patients, no abnormal uptake was observed. PET was positive in 12/14 patients with epithelioid peritoneal mesothelioma. Sensitivity, specificity and accuracy were respectively 86, 89 and 87%; the qualitative assessment was statistically different (P=0.0020, χ). Multicystic peritoneal mesothelioma histology was significantly associated with lower SUVmaxlesion (P=0.0061), SUVmaxlesion/liver (P=0.0025), Peritoneal Cancer Index, younger age, and it was observed only in women.Recurrence was observed on nine patients affected by epithelioid peritoneal mesothelioma, whereas no recurrences were observed among multicystic peritoneal mesothelioma patients. SUVmaxlesion (P=0.0278) and age (P=0.0241) were significantly associated with PFS in patients with epithelioid peritoneal mesothelioma. F-FDG-PET/ceCT showed significant differences between multicystic and epithelioid peritoneal mesothelioma, whereas SUVmaxlesion was associated with PFS in the latter. Although

  16. 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...

  17. CO adsorption on ionic Pt, Pd and Cu sites in Ce1−xMxO2−δ (M=Pt ...

    Indian Academy of Sciences (India)

    tuting noble metal ions in CeO2 and TiO2 supports on exhaust catalysis.12 When oxide supports are doped with lower valent noble metal ions, it creates oxygen vacancies in the support oxide to balance the overall charge. The study on CO oxidation by AuxCe1−xO2 catalyst (using standard DFT and DFT+U methods),.

  18. An innovative research reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Teruel, Federico E. [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Centro Atomico Bariloche, CNEA, Bariloche 8400, Rio Negro (Argentina)], E-mail: teruel@cab.cnea.gov.ar; Rizwan-uddin [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States)

    2009-02-15

    A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < {theta} < {pi}) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 x 10{sup 14} n cm{sup -2} s{sup -1}, 3.0 x 10{sup 14} n cm{sup -2} s{sup -1}, and 2.0 x 10{sup 14} n cm{sup -2} s{sup -1}, respectively. Heat deposition in the cladding, coolant and fuel material is

  19. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  20. Projected shell model study of neutron- deficient 122Ce

    Indian Academy of Sciences (India)

    Recently, Smith et al [1] have communicated the existence of neutron-deficient 122Ce and its excited states have been reported up to spin 14¯h. The band has been as- signed to 122Ce by detecting γ-rays in coincidence with evaporated charged particles and neutrons. The nucleus is believed to have large ground state ...

  1. Facile hydrothermal synthesis of CeO2 nanopebbles

    Indian Academy of Sciences (India)

    Administrator

    XRD shows the formation of cubic fluorite CeO2 and the average particle size estimated from the Scherrer formula was found to be 6.69 nm. X-ray absorption spectrum of CeO2 nanopebbles exhibits two .... state.26 In our case, the valence of cerium ions will be mainly 4+ state, because the estimated particle size of.

  2. Three-Dimensional Structure of CeO2 Nanocrystals

    DEFF Research Database (Denmark)

    Tan, Joyce Pei Ying; Tan, Hui Ru; Boothroyd, Chris

    2011-01-01

    truncated octahedral CeO2 particles through a lattice matched interface generated irregular compressed truncated octahedral CeO2 nanoparticles. The formation of this irregular shape is attributed to the lower surface diffusion and slow incorporation of atoms on surfaces by step attachment of the fused...

  3. 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 :

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

    International Nuclear Information System (INIS)

    Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Karol, R; Powell, R W

    1979-02-01

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

  6. Thermal stratification of sodium in the BN 600 reactor

    International Nuclear Information System (INIS)

    Obmelukhin, J.A.; Obukhov, P.I.; Rinejskij, A.A.; Sobolev, V.A.; Sherbakov, S.I.

    1983-01-01

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

  7. Low temperature synthesis of fluorite-type Ce-based oxides of ...

    Indian Academy of Sciences (India)

    M MALATHI

    type) Ce-based oxides of compositions Ln2Ce2O7 (Ln = Pr, Nd and Eu). F-type oxides can be considered as multi-functional materials due to the deficiency of anions. Pr2Ce2O7 (PCO), Nd2Ce2O7 (NCO) and Eu2Ce2O7 (ECO) were prepared ...

  8. Multimedia Experience on Web-Connected CE Devices

    Science.gov (United States)

    Tretter, Dan; Liu, Jerry; Zhang, Xuemei; Gao, Yuli; Atkins, Brian; Chao, Hui; Xiao, Jun; Wu, Peng; Lin, Qian

    Consumer electronics (CE) are changing from stand-alone single-function devices to products with increasing connectivity, convergence of functionality, and a focus on customer experience. We discuss the features that characterize the new generation of CE and illustrate this new paradigm through an examination of how web services can be integrated with CE products to deliver an improved user experience. In particular, we focus on one aspect of the CE segment, digital photography. We introduce AutoPhotobook, an automatic photobook creation service and provide a detailed look at how it addresses the complexity of photobook authoring through a portfolio of automatic photo analysis and composition technologies. We then show how this collection of technologies is integrated into a larger ecosystem with other web services and web-connected CE devices to deliver an enhanced user experience.

  9. 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.

  10. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    The invention deals with disengaging the coupling of a reactor coolant pump of a nuclear reactor feeding pressurized coolant. The disengaging coupling has two parts joined by bolts, at least one of them containing a driving agent within a bore. This is provided with a speed-depending ignition device in such manner that, if the critical speed is reached, the driving charge is ignited and the coupling is disengaged by destroying the bolts. (UWI) [de

  11. Neutron fluxes in test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  12. 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.

  13. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

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

  14. 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

  15. Tunable multicolor and white luminescence in Tb3+/Dy3+/Mn2+ doped CePO4 via energy transfer

    International Nuclear Information System (INIS)

    Yi, Zhigao; Lu, Wei; Zeng, Songjun; Wang, Haibo; Rao, Ling; Li, Zheng

    2015-01-01

    Highlights: • Tb 3+ /Dy 3+ /Mn 2+ doped CePO 4 DCNPs were prepared by a one-pot hydrothermal process. • The PL properties and the ET mechanism of these DCNPs were investigated in detail. • These DCNPs exhibit tunable multi-color output under UV excitation. • Intense white emissions can be realized by singly doping Dy 3+ and Mn 2+ in CePO 4 host. - Abstract: In this paper, a series of Tb 3+ /Dy 3+ /Mn 2+ doped CePO 4 downconversion nanophosphors (DCNPs) were prepared by a one-pot hydrothermal process. The obtained DCNPs presented monoclinic and hexagonal phase structure with wire-like shape. The photoluminescence (PL) properties and the energy transfer (ET) mechanism of these DCNPs were investigated in detail. The ET mechanism of Ce 3+ /Tb 3+ in CePO 4 host was calculated by means of concentration quenching and spectral overlapping, and calculation results revealed that dipole–dipole interactions should be more responsible. The maximum value of ET efficiency was measured to 87.4% for Tb 3+ doped CePO 4 system. In addition, owing to the efficient ET between Ce 3+ and Tb 3+ /Dy 3+ /Mn 2+ , these as-prepared DCNPs exhibit tunable multi-color output under ultra-violet (UV) light excitation. More importantly, the intense cold and warm white emissions can be realized by singly doping 2%Dy 3+ and 20%Mn 2+ in CePO 4 host under UV irradiation, respectively. The corresponding CIE 1931 coordinates were calculated to be (0.30, 0.30) and (0.30, 0.32), respectively, which are closed to the standard white emission (0.33, 0.33). These findings demonstrate the efficient white light emission by singly doped Dy 3+ or Mn 2+ in CePO 4 system for the first time, which is different from commonly used co-doped or tri-doped system. The multicolor tuning and white emission make these Tb 3+ /Dy 3+ /Mn 2+ doped CePO 4 DCNPs potential phosphors in the fields of displays, lighting, and field-emission devices

  16. Discussion about design basis flood of site of research reactors by river

    International Nuclear Information System (INIS)

    Rong Feng; Zhao Jianjun; Du Qiaomin; Zhang Lingyan

    2006-01-01

    This paper presents the well-defined standard in relation to design the basis flood of the sites of research reactors by river. It is based on the concept of some relational standards, analysis of hydrological calculation technology and methods, and analysis of accident dangerous degrees of research reactor, as well as in combination with the engineering practices. The flood preventing standard for research reactors with higher power should be the same with that of the nuclear power plants. (authors)

  17. PNAE and the Mandala System in the municipality of MAURITI/CE (Brazil

    Directory of Open Access Journals (Sweden)

    Tiago Cartaxo de Lucena

    2017-12-01

    Full Text Available The National School Feeding Program (PNAE in Mauriti/CE (Brazil has a positive impact on family farming, as it encourages and promotes the commercialization of its products. The goal of the program is to benefit students in basic education with healthy and nutritious food. It is worth mentioning that the PNAE is interconnected to the mandala system, whose principle is environmental sustainability and the improvement of the food standard, consequently improving the income and quality of life of the population. This study sought to analyze data from the PNAE in Mauriti/CE family agriculture secretariat from 2010 to 2014. The methodology used is descriptive statistics. The products of the PNAE produced by the mandala system had the highlight of the year 2012, the other years there was a reduction. However, with a view to strengthening family farming and reducing rural exodus, it is of great importance that this program is expanded.

  18. The mixed-valence state of Ce in the hexagonal CeNi sub 4 B compound

    CERN Document Server

    Tolinski, T; Pugaczowa-Michalska, M; Chelkowska, G

    2003-01-01

    Measurements of the magnetic susceptibility chi, x-ray photoemission spectra (XPS), electrical resistivity rho and electronic structure calculations for CeNi sub 4 B are reported. In the paramagnetic region, CeNi sub 4 B follows the Curie-Weiss law with mu sub e sub f sub f = 0.52 mu sub B /fu and theta -10.7 K. The effective magnetic moment is lower than the free Ce sup 3 sup + -ion value. The Ce(3d) XPS spectra have confirmed the mixed-valence state of Ce ions in CeNi sub 4 B. The f occupancy, n sub f , and the coupling DELTA between the f level and the conduction states were derived to be about 0.83 and 85 meV, respectively. Both susceptibility data and XPS spectra show that Ce ions in CeNi sub 4 B are in the intermediate-valence state. At low temperatures (below 12 K), the magnetic contribution to the electrical resistivity reveals a logarithmic slope characteristic of Kondo-like systems.

  19. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Preda, M.; Carabulea, A.

    2008-01-01

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

  20. Possible Frustration Effects on a New Antiferromagnetic Compound Ce6Pd13Zn4 with the Octahedral Ce Sublattice

    Science.gov (United States)

    Matsuoka, Eiichi; Oshima, Akihiro; Sugawara, Hitoshi; Sakurai, Takahiro; Ohta, Hitoshi

    2018-01-01

    Magnetization, specific heat, and electrical resistivity measurements have been performed on polycrystalline samples of a new cubic compound, Ce6Pd13Zn4. This compound exhibits metallic behavior and is classified as a Kondo-lattice system. The trivalent Ce ions are responsible for the antiferromagnetic transition at TN = 3.3 K and the phase transition at T'N = 1.3 K with the formation of superzone gaps. The increase in magnetic susceptibilities below TN and the considerably large value of the specific heat divided by temperature (1.25 J·Ce-mol-1·K-2) imply the existence of non-ordered Ce magnetic moments due to the geometrical frustration on the octahedral Ce sublattice.

  1. 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)

  2. Monte Carlo Simulation of Response Function for LaBr3(Ce) detector and Its Internal-Activity

    International Nuclear Information System (INIS)

    Negm, Hani; Omer, Mohamed; Zen, Heishun; Daito, Izuru; Kii, Toshiteru; Masuda, Kai; Hori, Toshitada; Ohgaki, Hideaki; Hajima, Ryoichi; Hayakawa, Takehito; Shizuma, Toshiyuki; Kikuzawa, Nobuhiro; Toyokawa, Hiroyuki

    2013-06-01

    Lanthanum halide scintillator LaBr 3 (Ce) could potentially improve the performance of spectroscopic systems because of their fine energy resolution in comparison with traditional scintillator materials and a good linearity of the energy response. Monte Carlo simulation of a detector response function presents a very challenging problem especially in 2 MeV region where internal radioisotopes inside the LaBr 3 (Ce) crystal make the energy response complex structure. The detector response functions of LaBr 3 (Ce) (38.1 mm x 76.2 mm cylindrical) have been calculated for standard γ-ray sources and internal sources using simulation toolkit GEANT4. The internal activity was investigated in details by comparison simulation results with the experimental. As a result we can reproduce the experimental spectra, however, there is still a few missing sources around 2 MeV region. (authors)

  3. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    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 /sup 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/sup -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.

  4. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

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

  5. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

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

  6. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

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

  7. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

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

  8. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  9. Annual progress report of the University of Florida Training Reactor September 1, 1979-August 31, 1980

    International Nuclear Information System (INIS)

    Diaz, N.J.

    1980-11-01

    Reported are: reactor operation, modifications, maintenance and tests, changes to technical specifications and standard operating procedures, radioactive releases and environmental surveillance, and training utilization

  10. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  11. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  12. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  13. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

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

  14. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

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

  15. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

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

  16. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

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

  17. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

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

  18. Power reactor embrittlement data base

    International Nuclear Information System (INIS)

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

    1989-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs

  19. 40 CFR 63.1407 - Non-reactor batch process vent provisions.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

  20. 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.

  1. New reactor concepts

    International Nuclear Information System (INIS)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

    1998-11-01

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

  2. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  3. Multi purpose research reactor

    International Nuclear Information System (INIS)

    Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

    2006-01-01

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor

  4. Trench reactor: an overview

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  5. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

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

  6. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  7. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

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

  8. Quality assurance plan, Westinghouse Water Reactor Divisions

    Energy Technology Data Exchange (ETDEWEB)

    1976-03-01

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

  9. Symmetries applied to reactor calculations

    International Nuclear Information System (INIS)

    Makai, M.

    1982-03-01

    Three problems of a reactor-calculational model are discussed with the help of symmetry considerations. 1/ A coarse mesh method applicable to any geometry is derived. It is shown that the coarse mesh solution can be constructed from a few standard boundary value problems. 2/ A second stage homogenization method is given based on the Bloch theorem. This ensures the continuity of the current and the flux at the boundary. 3/ The validity of the micro-macro separation is shown for heterogeneous lattices. A formula for the neutron density is derived for cell homogenization. (author)

  10. FCRD Advanced Reactor (Transmutation) Fuels Handbook

    Energy Technology Data Exchange (ETDEWEB)

    Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, important to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and

  11. FCRD Advanced Reactor (Transmutation) Fuels Handbook

    International Nuclear Information System (INIS)

    Janney, Dawn Elizabeth; Papesch, Cynthia Ann

    2016-01-01

    Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, important to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and

  12. Reactor feedwater facility

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Tadashi; Kinoshita, Shoichiro; Akatsu, Jun-ichi

    1996-04-30

    In a reactor feedwater facility in which one stand-by system and at least three ordinary systems are disposed in parallel, each of the feedwater pumps is driven by an electromotor, and has substantially the same capacity. At least two systems among the ordinary systems have a pump rotation number variable means. Since the volume of each of the feedwater pump of each system is determined substantially equal, standardization is enabled to facilitate the production. While the number of electromotors is increased, since they are driven by electromotors, turbines, steam pipelines and valves for driving feed water pumps can be eliminated. Therefore, the feedwater pumps can be disposed to a region of low radiation dose being separated from a main turbine and a main condensator, to improve the degree of freedom in view of the installation. In addition, accessibility to equipments during operation is improved to improve the maintenance of feed water facilities. The number of parts for equipments can be reduced compared with that in a turbine-driving system thereby capable of reducing the operation amount for the maintenance and inspection. (N.H.)

  13. Results of the Nucifer reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

    2016-07-01

    Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

  14. MOX in reactors: present and future

    International Nuclear Information System (INIS)

    Arslan, Marc; Gros, Jean Pierre; Niquille, Aurelie; Marincic, Alexis

    2010-01-01

    In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR TM or ATMEA TM designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR TM reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR TM can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO 2 , ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)

  15. Luncheon address: Development of the CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bain, A.S. [Atomic Energy of Canada Limited (Canada)

    1997-07-01

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

  16. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    Bain, A.S.

    1997-01-01

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

  17. Commissioning of research reactors. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

  18. Reactor core monitor for nuclear reactor

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    The device of the present invention provides a various information of a wide adaptability, such as a power distribution, to an operator by determining a reactor core performance of the reactor by a performance calculation with improved accuracy. That is, a calculation means determines a neutron flux distribution of the reactor and coolant temperature based on the neutron flux distribution. A measuring means measures a cooled temperature of a reactor core inlet and a temperature at the exit of a fuel assembly. The result of coolant temperature by the measuring means and the result of the calculation by the calculation means are compared. The result of the calculation for the neutron flux distribution obtained by the calculation means is corrected based on the result of the comparison. The calculation means introduces calculation at higher accuracy by adopting two-dimensional balance in the fuel assembly. Further, a more accurate three-dimensional neutron diffusion calculation model is introduced in an on-line computer. Then, the accuracy of the calculation for the neutron flux distribution, power distribution, temperature distribution, etc. is improved. In view of the above, adaptability of a reactor core monitor is widened. (I.S.)

  19. 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

  20. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  1. A new adsorbent of a Ce ion-implanted metal-organic framework (MIL-96) with high-efficiency Ce utilization for removing fluoride from water.

    Science.gov (United States)

    Yang, Xuan; Deng, Shuangshuang; Peng, Fumin; Luo, Tao

    2017-02-14

    A novel Ce(iii) ion-implanted aluminum-trimesic metal-organic framework (Ce-MIL-96) was synthesized for the first time via alcohol-solvent incipient wetness impregnation. Compared to previously reported Ce-contained adsorbents, the fluoride adsorption performance of the new ion-implanted metal-organic framework demonstrated much higher adsorption capacity and more efficient regeneration of Ce. In a wide pH range of 3 to 10, Ce-MIL-96 maintained constant adsorption performance for fluoride, and the residual Ce and Al in the treated solution were below the safe limits in drinking water. The maximum adsorption capacity of Ce-MIL-96 was 38.65 mg g -1 at 298 K. Excluding the contribution of MIL-96, the maximum adsorption capacity of Ce ions was 269.75 mg g -1 , which demonstrated that the service efficiency of cerium in Ce-MIL-96 is about 6 times that in Ce 2 O 3 , nearly 10 times that in Ce-mZrp, and double that in Mn-Ce oxides. There was no significant influence on fluoride removal by Ce-MIL-96 due to the presence of chloride, nitrate, sulfate, bicarbonate or phosphate. Moreover, the adsorption capacity of Ce-MIL-96 remained at more than 70% after nine cycles of adsorption-desorption. Due to this excellent adsorption performance and its regeneration properties, Ce-MIL-96 is a promising adsorbent for the removal of fluoride from groundwater.

  2. 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

  3. Department of Energy Nuclear Energy Standards Program

    International Nuclear Information System (INIS)

    Silver, E.G.

    1980-01-01

    The policy with respect to the development and use of standards in the Department of Energy (DOE) programs concerned with maintaining and developing the nuclear option for the civilian sector (both in the form of the currently used light water reactors and for advanced concepts including the Liquid Metal Fast Breeder Reactor), is embodied in a Nuclear Standards Policy, issued in 1978, whose perspectives and philosophy are discussed

  4. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  5. New research reactor for Australia

    International Nuclear Information System (INIS)

    Miller, R.

    1992-01-01

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

  6. 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)

  7. Negative Ce anomalies in Mn oxides: The role of Ce4+ mobility during water-mineral interaction

    Science.gov (United States)

    Loges, Anselm; Wagner, Thomas; Barth, Matthias; Bau, Michael; Göb, Susanne; Markl, Gregor

    2012-06-01

    We present one of the very rare natural examples of extremely negative Ce anomalies (up to 4 orders of magnitude) in manganese oxides, caused by higher mobility of Ce4+ compared to REE3+ in an aquatic environment. The young secondary Mn oxides formed together with fluorites and goethites during water-mineral interaction in a hydrothermal fluorite vein. Our findings are in contrast to the oxidative scavenging of Ce, which is commonly observed in Mn oxides. Comparison of REE patterns from modern mine waters with primary and secondary minerals demonstrates that this cannot be solely explained as a source-related feature or by immobilization of Ce, but must at least partially be the result of preferential mobilization of Ce4+ compared to REE3+. We propose that this very unusual behavior is due to strong complexation of Ce4+, most likely by siderophores or similar organic molecules, based on published complex formation constants. The presence of dissolved organic carbon (DOC) even in water samples outflowing at the deepest mine level lends support to this model. Recent experimental studies have also demonstrated that negative Ce anomalies develop in biogenic Mn oxides, but this effect has not yet been reported from a natural environment. Our findings emphasize the relevance of experimental results for natural systems and have considerable implications for the assessment of the mobility of tetravalent actinide elements (especially Pu4+) in earth surface environments or potential high-level permanent repositories.

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

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

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

  9. Panel plenary session: Status and future needs in the field of reactor safety research

    International Nuclear Information System (INIS)

    Finzi, S.; Cicognani, G.; Heusener, G.; Geijzers, H.F.G.; Alonso-Santos, A.; Holtbecker, H.F.

    1990-01-01

    Status and future needs in the field of reactor safety research. Overviews are given of the nuclear programme in France and the Netherlands. Spanish and Italian reactor safety research both current and for the future is outlined. LWR safety and the continuation of the establishment of safety standards particularly for LMFBR reactors is discussed. The new framework for the research in reactor safety by the Commission of the European Communities for 1990-1994 is outlined. The discussion which followed is reported. (UK)

  10. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  11. A nuclear power reactor

    International Nuclear Information System (INIS)

    Borrman, B.E.; Broden, P.; Lundin, N.

    1979-12-01

    The invention consists of shock absorbing support beams fastened to the underside of the reactor tank lid of a BWR type reactor, whose purpose is to provide support to the steam separator and dryer unit against accelerations due to earthquakes, without causing undue thermal stresses in the unit due to differential expansion. (J.I.W.)

  12. Reactor cost driving items

    International Nuclear Information System (INIS)

    Spears, W.R.

    1987-01-01

    Assuming that the design solutions presently perceived for NET can be extrapolated for use in a power reactor, and using costing experience with present day fusion experiments and with fission power plants, the major components of the cost of a tokamak fusion power reactor are described. The analysis shows the emphasis worth placing on various areas of plant design to reduce costs

  13. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

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

  14. CAREM 25 nuclear reactor

    International Nuclear Information System (INIS)

    Rossini, A.A.; Ordonez, J.P.; Rajoy, J.E.; Durione, C.

    1990-01-01

    This work describes the CAREM project reactor, its design philosophy, its main characteristics and its advantages with respect to similar reactors. The main objective is to use the nuclear energy at lower costs than those applied up to now. (Author) [es

  15. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

    Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

  16. International thermal reactor development

    International Nuclear Information System (INIS)

    Zebroski, E.L.

    1977-01-01

    The worldwide development of nuclear power plants is reviewed. Charts are presented which show the commitment to light-water reactor capacity construction with breakdown by region and country. Additional charts show the major nuclear research centers which have substantial scope in light water reactor development and extensive international activities

  17. The fusion reactor

    International Nuclear Information System (INIS)

    Brennan, M.H.

    1974-01-01

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

  18. Advanced converter reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-01-01

    Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U 3 O 8 at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals

  19. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

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

  20. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  1. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Lineberry, M.J.

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs

  2. Nuclear reactor instrumentation method

    International Nuclear Information System (INIS)

    Handa, Hiroyuki; Hayashi, Katsumi; Nemesawa, Shigeki; Nemoto, Yuji; Ohashi, Masahisa.

    1993-01-01

    The present invention can appropriately monitor the state of a reactor core in an FBR type reactor which has a system of storing spent fuel assemblies in a reactor container while reducing the weight and making the structure compact in the reactor. That is, a fuel assembly having a shield lacking portion in upper axial shields is disposed. The shield lacking portion defines neutrons' leaking path from the reactor core. The leakage of neutrons from the path is detected by a neutron monitor disposed just above the fuel assembly. With such a constitution, influence of neutrons from stored spent fuel assemblies disposed to the out side of the radial shields can be reduced by a shielding effect of the existent radial shields around the reactor core. Further, if a shield lacking portion is locally disposed in the region of the upper axial shields just below the neutron monitor, neutrons from the reactor core can be monitored while suppressing excessive neutron leakage. As a result, it is unnecessary to dispose shields on the outer side of the spent fuel assembly disposed in the reactor core. (I.S.)

  3. Reactor building for a nuclear reactor

    International Nuclear Information System (INIS)

    Haidlen, F.

    1976-01-01

    The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

  4. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  5. The System 80+ Standard Plant Information Management System

    International Nuclear Information System (INIS)

    Turk, R.S.; Bryan, R.E.

    1998-01-01

    Historically, electric nuclear power plant owners, following the completion of construction and startup, have been left with a mountain of hard-copy documents and drawings. Hundreds of thousands of hours are spent searching for relevant documents and, in most cases, the documents found require many other documents and drawings to fully understand the design basis. All too often the information is incomplete, and eventually becomes obsolete. In the U.S., utilities spend millions of dollars to discover design basis information and update as-built data for each plant. This information must then be stored in an easily accessed usable form to assist satisfy regulatory requirements and to improve plant operating efficiency. ABB Combustion Engineering Nuclear Systems (ABB-CE) has an active program to develop a state-of-the-art Plant Information Management System (IMS) for its advanced light water reactor, the System 80+TM Standard Plant Design. This program is supported by ABB's Product Data Management (PDM) and Computer Aided Engineering (CAE) efforts world wide. This paper describes the System 80+ plant IMS and how it will be used during the entire life cycle of the plant. (author)

  6. The System 80+ Standard Plant Information Management System

    Energy Technology Data Exchange (ETDEWEB)

    Turk, R.S.; Bryan, R.E. [ABB Combuions Engineering Nuclear Systems (United States)

    1998-07-01

    Historically, electric nuclear power plant owners, following the completion of construction and startup, have been left with a mountain of hard-copy documents and drawings. Hundreds of thousands of hours are spent searching for relevant documents and, in most cases, the documents found require many other documents and drawings to fully understand the design basis. All too often the information is incomplete, and eventually becomes obsolete. In the U.S., utilities spend millions of dollars to discover design basis information and update as-built data for each plant. This information must then be stored in an easily accessed usable form to assist satisfy regulatory requirements and to improve plant operating efficiency. ABB Combustion Engineering Nuclear Systems (ABB-CE) has an active program to develop a state-of-the-art Plant Information Management System (IMS) for its advanced light water reactor, the System 80+TM Standard Plant Design. This program is supported by ABB's Product Data Management (PDM) and Computer Aided Engineering (CAE) efforts world wide. This paper describes the System 80+ plant IMS and how it will be used during the entire life cycle of the plant. (author)

  7. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

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

  8. Iris reactor conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V. [Westinghouse Electric Comp., Pittsburgh, PA (United States); Galvin, M.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Lombardi, C.V.; Maldari, F.; Ricotti, M.E. [Politecnico di Milano, Milan (Italy); Cinotti, L. [Ansaldo SpA, Genoa (Italy)

    2001-07-01

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

  9. Reactor control device

    International Nuclear Information System (INIS)

    Fukami, Haruo; Morimoto, Yoshinori.

    1981-01-01

    Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

  10. Mirror reactor surface study

    International Nuclear Information System (INIS)

    Hunt, A.L.; Damm, C.C.; Futch, A.H.; Hiskes, J.R.; Meisenheimer, R.G.; Moir, R.W.; Simonen, T.C.; Stallard, B.W.; Taylor, C.E.

    1976-01-01

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

  11. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

  12. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  13. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

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

  14. Hydrogenation of furfural at the dynamic Cu surface of CuOCeO2/Al2O3 in vapor phase packed bed reactor

    Science.gov (United States)

    The hydrogenation of furfural to furfuryl alcohol over a CuOCeO2/'-Al2O3 catalyst in a flow reactor is reported. The catalyst was prepared by the wet impregnation of Cu onto a CeO2/'-Al2O3 precursor. The calcined catalyst was then treated with HNO3 to remove surface CuO resulting in a mixed CuCe oxi...

  15. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1984-08-01

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

  16. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

    1977-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

  17. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1976-09-01

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

  18. 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.

  19. 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)

  20. Temperature-independent photoluminescence response in ZnO:Ce ...

    Indian Academy of Sciences (India)

    2017-12-06

    Ce3+ nanophosphors prepared by combustion method and solid-state reaction method is presented in this study. The powder XRD exhibits hexagonal wurtzite phase and crystallite size falls in the nanometre range. The optical ...

  1. Uued IT-mõtted CeBITilt / Kai Simson

    Index Scriptorium Estoniae

    Simson, Kai

    2006-01-01

    Riigi Infosüsteemide Arenduskeskuse (RIA) pressiesindaja Rica Semjonova vahendab Eesti stendistide muljeid Hannoveris toimunud IT-messilt CeBIT, kus Eestit tutvustasid nii riigiametid kui eraettevõtted

  2. Safety considerations for research reactors in extended shutdown

    International Nuclear Information System (INIS)

    2004-01-01

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

  3. Multi-purpose reactor

    International Nuclear Information System (INIS)

    1991-05-01

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  4. Hydrogen-bridge Si(μ-H)3CeH and inserted H3SiCeH molecules: Matrix infrared spectra and DFT calculations for reaction products of silane with Ce atoms

    Science.gov (United States)

    Xu, Bing; Shi, Peipei; Huang, Tengfei; Wang, Xuefeng

    2017-10-01

    Reactions of laser-ablated cerium atoms with silane were investigated by matrix isolation infrared spectroscopy and theoretical calculations. The reaction products, Si(μ-H)3CeH, H3SiCeH, H2Si(μ-H)CeH and HSi(μ-H)2CeH were identified on the basis of the SiD4 isotopic substitutions and DFT frequency calculations. In the solid argon or krypton matrix, the inserted H3SiCeH molecule was observed as initial product on deposition, which rearranged to hydrogen bridge species Si(μ-H)3CeH on follow-up annealing through H2Si(μ-H)CeH and HSi(μ-H)2CeH species. The Sisbnd Hsbnd Ce hydrogen bridge was investigated by NBO and ELF analysis. Calculation suggested that in Si(μ-H)3CeH molecule Ce atom donated one electron to Si atom, resulting in electron-rich SiH3 subunit, which was confirmed by ESP and AIM analysis. The increased basicity of Sisbnd H bond facilitates the formation of hydrogen bridge bond between Si and Ce. For comparison only insertion H3CCeH structure was obtained from the reaction of Ce atoms with CH4.

  5. Safety Management at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Zarina Masood; Ahmad Nabil Abdul Rahim

    2011-01-01

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

  6. Seamless Ligation Cloning Extract (SLiCE) Cloning Method

    OpenAIRE

    Zhang, Yongwei; Werling, Uwe; Edelmann, Winfried

    2014-01-01

    SLiCE (Seamless Ligation Cloning Extract) is a novel cloning method that utilizes easy to generate bacterial cell extracts to assemble multiple DNA fragments into recombinant DNA molecules in a single in vitro recombination reaction. SLiCE overcomes the sequence limitations of traditional cloning methods, facilitates seamless cloning by recombining short end homologies (15–52 bp) with or without flanking heterologous sequences and provides an effective strategy for directional subcloning of D...

  7. Description and Application of the CE-SE Scheme

    Directory of Open Access Journals (Sweden)

    Martin KOSÍK

    2010-12-01

    Full Text Available The CE-SE scheme is new numerical methodology for conservation laws. It was developed by Dr.Chang of NASA Glenn Research Center and his collaborators. The 1D and 2D variant of CE-SE scheme for scalar convection and for Euler equations for incompressible flows is described. The CESE scheme is compared with classical schemes in 1D. The solution to inviscid incompressible flow in GAMM channel in 2D is presented.

  8. Dissolved cerium contributes to uptake of Ce in the presence of differently sized CeO2-nanoparticles by three crop plants.

    Science.gov (United States)

    Schwabe, Franziska; Tanner, Simon; Schulin, Rainer; Rotzetter, Aline; Stark, Wendelin; von Quadt, Albrecht; Nowack, Bernd

    2015-03-01

    We investigated the uptake of cerium (Ce) dioxide nanoparticles (NPs) by hydroponically grown wheat, pumpkin and sunflower plants. The presence of plant roots in nutrient solution led to a substantial increase in the dissolution of CeO2-NP compared to plant-free medium. Experiments with Zr/CeOx-NP revealed that Ce was not only taken up in the form of NPs, but simultaneously to a significant degree also as dissolved Ce(iii) ions, which then re-precipitated in the form of CeO2-NPs inside the leaves. The contribution of dissolved Ce uptake was particularly large for particles smaller than 10 nm due to their higher dissolution rate. Our data also indicate that the translocation of Ce resulting from NP-root-exposure is species dependent. When Ce was supplied as dissolved ions, sunflower had the highest capacity of Ce-ion accumulation inside the leaves, while there was no significant difference between pumpkin and wheat. We found no Ce translocation from roots into shoots when only NPs bigger than 20 nm were applied. This study highlights that plant root activity can have a significant impact on the dissolution of CeO2-NPs in soil solution and that uptake of dissolved Ce(iii) followed by re-precipitation needs to be considered as an important pathway in studies of CeO2-NP uptake by plants.

  9. Dynamic simulation of a pilot scale vacuum gas oil hydrocracking unit by the space-time CE/SE method

    Energy Technology Data Exchange (ETDEWEB)

    Sadighi, S.; Ahmad, A. [Institute of Hydrogen Economy, Universiti Teknologi Malaysia, Johor Bahru (Malaysia); Shirvani, M. [Faculty of Chemical Engineering, University of Science and Technology, Tehran (Iran, Islamic Republic of)

    2012-05-15

    This work introduces a modified space-time conservation element/solution element (CE/SE) method for the simulation of the dynamic behavior of a pilot-scale hydrocracking reactor. With this approach, a four-lump dynamic model including vacuum gas oil (VGO), middle distillate, naphtha and gas is solved. The proposed method is capable of handling the stiffness of the partial differential equations resulting from the hydrocracking reactions. To have a better judgment, the model is also solved by the finite difference method (FDM), and the results from both approaches are compared. Initially, the absolute average deviation of the cold dynamic simulation using the CE/SE approach is 8.98 %, which is better than that obtained using the FDM. Then, the stability analysis proves that for achieving an appropriate response from the dynamic model, the Courant number, which is a function of the time step size, mesh size and volume flow rate through the catalytic bed, should be less than 1. Finally, it is found that, following a careful selection of these parameters, the CE/SE solutions to the hydrocracking model can produce higher accuracy than the FDM results. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  10. CeO{sub 2} nanoparticles deposited on carbon nanotubes

    Energy Technology Data Exchange (ETDEWEB)

    Zheng Guobin; Nomiyama, Ryoma; Sano, Hideaki; Uchiyama, Yasuo, E-mail: gbzheng@nagasaki-u.ac.jp [Department of Materials Science and Engineering, Nagasaki University, Bunkyo machi 1-14, Nagasaki, 852-8521 (Japan)

    2011-10-29

    Carbon nanotubes (CNTs) with average diameter of 14 nm were modified by oxidation in nitric acid and followed by treatment in sodium dodecyl sulfate (SDS). XPS analysis verified the introduction of carboxyl group and SDS molecules onto the CNT surface. The CeO{sub 2} coatings on CNTs were synthesized using two different processes; chemical precipitation method and solvothermal method. TEM observation showed that the coatings consisted of CeO{sub 2} nanoparticles with size of several nanometers. It is believed that in chemical precipitation method, the surface modification with carboxyl group and SDS provided nucleation sites of CeO{sub 2}, thus facilitating further growth and attachment of CeO{sub 2} nanoparticles. In solvothermal method, due to the interaction between graphene surface of CNTs and solvent pyridine, CeO{sub 2} particles could form on the pristine CNTs, and the surface modification decreased the nucleation sites of CeO{sub 2} and thus led to thinner or incomplete coating.

  11. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1970-04-01

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

  12. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  13. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1993-09-01

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

  14. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    1990-09-01

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

  15. Ce O{sub 2}-Zr O{sub 2} tetragonal ceramics (Ce-TZP): mechanical properties

    Energy Technology Data Exchange (ETDEWEB)

    Nono, Maria do Carmo de Andrade [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Laboratorio Associado de Sensores e Materiais

    1994-12-31

    This work presents the development and the characterization of Ce O{sub 2}-stabilized tetragonal Zr O{sub 2} polycrystals (Ce-TZP ceramics), since it is considered candidate material for applications as structural high performance ceramics, an as substitute of some metallic materials. Ce-TZP ceramics attain remarkable increasing in strength and fracture toughness. Sintered ceramics were fabricated from mixtures of powders containing different Ce O{sub 2} content prepared by conventional mechanical technique. It were adopted the bending strength, Vickers hardness and fracture toughness techniques to the determination of the mechanical parameters. These results were discussed and compared to those published in international literature. (author) 11 refs., 10 figs.

  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. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

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

  19. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  20. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  1. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  2. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  3. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  4. Water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Donaldson, A.J.

    1989-01-01

    In order to reduce any loss of primary water coolant from around a reactor core of a water cooled nuclear reactor caused by any failure of a pressure vessel, an inner vessel is positioned within and spaced from the pressure vessel. The reactor core and main portion of the primary water coolant circuit and a heat exchanger are positioned within the inner vessel to maintain some primary water coolant around the reactor core and to allow residual decay heat to be removed from the reactor core by the heat exchanger. In the embodiment shown an aperture at the upper region of the inner vessel is dimensioned configured and arranged to prevent steam from a steam space of an integral pressurised water cooled nuclear reactor for a ship entering the main portion of the primary water coolant circuit in the inner vessel if the longitudinal axis of the nuclear reactor is displaced from its normal substantially vertical position to an abnormal position at an angle to the vertical direction. Shields are integral with the inner vessel. (author)

  5. Reactor monitoring device

    International Nuclear Information System (INIS)

    Kono, Shigehiro.

    1996-01-01

    The device of the present invention monitors the stability of a power of a BWR type reactor by using each of recycling flow rates in addition to a reactor core flow rate to improve monitoring accuracy. Namely, a set value registering means is disposed for registering reactor core flow rate set values corresponding to the number of recycling flow rates not reaching a reference value for each of the recycling flow rates. A reactor flow rate take-out means judges whether each of the recycling flow rates reaches the reference value or not. The set values of the set value registering means are taken out based on the number of each of the recycling flow rate signals not reaching the reference values. The taken out set value and calculated reactor core flow rate value are compared by an abnormal alarm means. When calculated value is smaller than the set value, abnormality is informed. The accuracy for the monitoring is improved by monitoring the reactor power by using each of recycling flow rates in addition to the reactor core flow rate. (I.S.)

  6. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  7. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  8. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  9. Reactor safety device

    International Nuclear Information System (INIS)

    Okada, Yasumasa.

    1987-01-01

    Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

  10. Interfacing HTCondor-CE with OpenStack

    Science.gov (United States)

    Bockelman, B.; Caballero Bejar, J.; Hover, J.

    2017-10-01

    Over the past few years, Grid Computing technologies have reached a high level of maturity. One key aspect of this success has been the development and adoption of newer Compute Elements to interface the external Grid users with local batch systems. These new Compute Elements allow for better handling of jobs requirements and a more precise management of diverse local resources. However, despite this level of maturity, the Grid Computing world is lacking diversity in local execution platforms. As Grid Computing technologies have historically been driven by the needs of the High Energy Physics community, most resource providers run the platform (operating system version and architecture) that best suits the needs of their particular users. In parallel, the development of virtualization and cloud technologies has accelerated recently, making available a variety of solutions, both commercial and academic, proprietary and open source. Virtualization facilitates performing computational tasks on platforms not available at most computing sites. This work attempts to join the technologies, allowing users to interact with computing sites through one of the standard Computing Elements, HTCondor-CE, but running their jobs within VMs on a local cloud platform, OpenStack, when needed. The system will re-route, in a transparent way, end user jobs into dynamically-launched VM worker nodes when they have requirements that cannot be satisfied by the static local batch system nodes. Also, once the automated mechanisms are in place, it becomes straightforward to allow an end user to invoke a custom Virtual Machine at the site. This will allow cloud resources to be used without requiring the user to establish a separate account. Both scenarios are described in this work.

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

    Czech Academy of Sciences Publication Activity Database

    Kubešová, Marie; Kučera, Jan

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Hart, R.S.

    1997-01-01

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

  13. Optical and scintillation properties of Ce-doped LuLiF{sub 4} with different Ce concentrations

    Energy Technology Data Exchange (ETDEWEB)

    Yanagida, Takayuki, E-mail: yanagida@lsse.kyutech.ac.jp [Kyushu Institute of Technology, 2-4 Hibikino, Wakamatsu-ku, Kitakyushu, Fukuoka 808-0196 (Japan); Fujimoto, Yutaka [Kyushu Institute of Technology, 2-4 Hibikino, Wakamatsu-ku, Kitakyushu, Fukuoka 808-0196 (Japan); Fukuda, Kentaro [Tokuyama Corporation, Shibuya 3-chome, Shibuya-ku, Tokyo 150-8383 Japan (Japan); Chani, Valery [Tohoku Univ., 2-1-1 Katahira, Sendai 980-8577 (Japan)

    2013-11-21

    The crystals of 0.1, 0.5, and 1 mol% Ce-doped LuLiF{sub 4} (Ce:LLF) grown by the micro-pulling down (μ-PD) method were examined for their optical and scintillation properties. Ce:LLF crystals had ∼80% transparency at wavelengths longer than 300 nm. In photoluminescence spectra, they demonstrated intense emission peaks at 310 and 330 nm with the quantum yield of 60–90%. Ce{sup 3+} 5d–4f emission peaks were also detected at similar wavelengths of 310 and 330 nm in the radioluminescence spectra obtained under X-ray excitation. According to pulse height spectra recorded under γ-ray irradiation, the absolute light yield of Ce 0.1, 0.5, and 1% were 3600±400, 3000±300, and 1700±200 ph/MeV, respectively. Decay time kinetics was also inspected using a pulse X-ray equipped streak camera system. The decay time components of Ce:LLF were ∼70 ns and ∼1 μs for all the samples.

  14. Optical and scintillation properties of Ce-doped LuLiF4 with different Ce concentrations

    International Nuclear Information System (INIS)

    Yanagida, Takayuki; Fujimoto, Yutaka; Fukuda, Kentaro; Chani, Valery

    2013-01-01

    The crystals of 0.1, 0.5, and 1 mol% Ce-doped LuLiF 4 (Ce:LLF) grown by the micro-pulling down (μ-PD) method were examined for their optical and scintillation properties. Ce:LLF crystals had ∼80% transparency at wavelengths longer than 300 nm. In photoluminescence spectra, they demonstrated intense emission peaks at 310 and 330 nm with the quantum yield of 60–90%. Ce 3+ 5d–4f emission peaks were also detected at similar wavelengths of 310 and 330 nm in the radioluminescence spectra obtained under X-ray excitation. According to pulse height spectra recorded under γ-ray irradiation, the absolute light yield of Ce 0.1, 0.5, and 1% were 3600±400, 3000±300, and 1700±200 ph/MeV, respectively. Decay time kinetics was also inspected using a pulse X-ray equipped streak camera system. The decay time components of Ce:LLF were ∼70 ns and ∼1 μs for all the samples

  15. Iodine chemistry in a reactor regulation

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  16. Inherently safe reactors

    International Nuclear Information System (INIS)

    Maartensson, Anders

    1992-01-01

    A rethinking of nuclear reactor safety has created proposals for new designs based on inherent and passive safety principles. Diverging interpretations of these concepts can be found. This article reviews the key features of proposed advanced power reactors. An evaluation is made of the degree of inherent safety for four different designs: the AP-600, the PIUS, the MHTGR and the PRISM. The inherent hazards of today's most common reactor principles are used as reference for the evaluation. It is concluded that claims for the new designs being inherently, naturally or passively safe are not substantiated by experience. (author)

  17. Reactor safety assessment system

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  18. Fast breeder reactors

    International Nuclear Information System (INIS)

    1978-01-01

    The subject of this invention is a liquid metal cooled nuclear reactor construction in which a concrete pit is lagged to protect it from the heat radiated from the reactor in normal operation but in which the efficiency of the lagging is reduced in case of emergency to allow the excess heat generated by the reactor to be dissipated throughout the pit. The lagging is in two layers, the first covering the internal surface of the pit wall is impermeable to the liquid metal, whilst the second layer over the first is permeable [fr

  19. Reactor shutdown device

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1983-01-01

    Purpose : To provide a reactor shutdown device suitable to the low temperature shutdown of a heavy water-moderated type nuclear reactor and capable of ensuring an adequate shutdown margin. Constitution : Xenon reactivity is calculated based on the detection signals for reactor neutrons, the temperature reactivity is calculated based on the temperature of the moderators and of the coolants and, further, poisons in the moderators are detected. Injection amount of the poisons is calculated based on the result of the calculation and the detection, and the calculated amount of poisons is injected into the moderators. (Kamimura, M.)

  20. Reactor flux calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lhuillier, D. [Commissariat à l' Énergie Atomique et aux Énergies Alternatives, Centre de Saclay, IRFU/SPhN, 91191 Gif-sur-Yvette (France)

    2013-02-15

    The status of the prediction of reactor anti-neutrino spectra is presented. The most accurate method is still the conversion of total β spectra of fissionning isotopes as measured at research reactors. Recent re-evaluations of the conversion process led to an increased predicted flux by few percent and were at the origin of the so-called reactor anomaly. The up to date predictions are presented with their main sources of error. Perspectives are given on the complementary ab-initio predictions and upcoming experimental cross-checks of the predicted spectrum shape.