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Sample records for general electric standard reactor

  1. 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

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

    International Nuclear Information System (INIS)

    Bottimore, R.R.

    1980-12-01

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

  3. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

  4. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  5. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  6. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  7. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  8. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  9. Generic risk insights for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    Travis, R.; Taylor, J.; Chung, J.

    1991-05-01

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs

  10. 76 FR 70166 - Electrical Standards for Construction and General Industry; Extension of the Office of Management...

    Science.gov (United States)

    2011-11-10

    ...] Electrical Standards for Construction and General Industry; Extension of the Office of Management and Budget... contained in the Electrical Standards for Construction (29 CFR part 1926, Subpart K) and for General... maintenance of electric utilization equipment that prevent death and serious injuries among construction and...

  11. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

  12. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  13. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

  14. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  15. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

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

    Science.gov (United States)

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

    2017-06-28

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

  17. NRC [Nuclear Regulatory Commission] staff evaluation of the General Electric Company Nuclear Reactor Study (''Reed Report'')

    International Nuclear Information System (INIS)

    1987-07-01

    In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, also referred to as ''the Reed Report,'' an internal product-improvement study. GE considered the document ''proprietary'' and thus, under the regulations of the Nuclear Regulatory Commission (NRC), exempt from mandatory public disclosure. Nonetheless, members of the NRC staff reviewed the document in 1976 and determined that it did not raise any significant new safety issues. The staff also reached the same conclusion in subsequent reviews. However, in response to recent inquiries about the report, the staff reevaluated the Reed Report from a 1987 perspective. This re-evaluation, documented in this staff report, concluded that: (1) there are no issues raised in the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR); (2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and (3) although certain issues addressed by the Reed Report are still being studied by the NRC and the industry, there is no basis for suspending licensing and operation of GE BWR plants while these issues are being resolved

  18. Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

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

    1993-08-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set

  19. Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

    International Nuclear Information System (INIS)

    D'Ardenne, W.H.; Sherwood, G.G.

    1985-01-01

    The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)

  20. Standard Technical Specifications General Electric plants, BWR/4:Bases (Sections 3.4-3.10). Volume 3, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the specifications for all chapters and sections of the improved STS. Volume 2 contains he Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. This document, Volume 3, contains the Bases for Sections 3.4-3.10 of the improved STS

  1. Standard Technical Specifications General Electric plants, BWR/4: Bases (Sections 2.0-3.3). Volume 2, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved ST or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume I contains the Specifications for all chapters and sections of the improved STS. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4-3.10 of the improved STS

  2. Safety Evaluation Report related to the renewal of the operating license for the General Electric-Nuclear Test Reactor (GE-NTR) (Docket No. 50-73)

    International Nuclear Information System (INIS)

    1984-09-01

    This Safety Evaluation Report for the application filed by the General Electric Company (GE) for a renewal license number R-33 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by GE and is located in Pleasanton, California. The staff concludes that the reactor can continue to be operated by GE without endangering the health and safety of the public

  3. TRAC development at General Electric

    International Nuclear Information System (INIS)

    Andersen, J.G.M.; Shaug, J.C.; Shiralkar, B.S.

    1987-01-01

    TRAC is a computer code for transient analysis of light water reactors. The BWR version of TRAC has been developed as a result of a close cooperation between General Electric Company and Idaho National Engineering Laboratory. Up through 1985 the development work at General Electric was jointly funded by General Electric, the Nuclear Regulatory Commission and Electric Power Research Institute under the Refill-Reflood and FIST programs. At INEL (which has the main responsibility for the NRC version of TRAC-BWR) this work has led to the development of TRACBD1 and TRACBF1, while at GE, TRACB04 was the final product of the Refill-Reflood and FIST programs. TRAC development has continued at General Electric after the completion of these programs with the evolution of the TRACG code. The purpose of the paper is to describe this work. The TRAC development at General Electric can be divided into two main categories: extended benchmark capability and improved user convenience

  4. 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...

  5. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  6. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  7. Consideration of severe accident issues for the General Electric BWR standard plant: Chapter 10

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

    In early 1982, the U.S. Nuclear Regulatory Commission (NRC) proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. GE provided appendices to the licensing documentation of its standard plant design, GESSAR II, which address severe accidents for the GE BWR/6 Mark III 238 nuclear island design. The GE submittals discuss the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at Three Mile Island (TMI), is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

  8. Consideration of severe accident issues for the general electric BWR standard plant a status report

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

    In early 1982 the U.S. NRC proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. This paper, GE's submission, discusses the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at TMI, is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

  9. 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

  10. Nozzle for electric dispersion reactor

    Science.gov (United States)

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1995-11-07

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

  11. 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

  12. TU electric reactor model verification

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1989-01-01

    Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

  13. 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.

  14. 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

  15. Reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

  16. Reactors for nuclear electric propulsion

    International Nuclear Information System (INIS)

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades

  17. 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

  18. The national standards program for research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1977-01-01

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

  19. 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.

  20. 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

  1. General Electric Nuclear Energy

    International Nuclear Information System (INIS)

    2001-01-01

    The ESBWR is a 1380 MWe boiling water reactor with improved operating safety margins and passive safety systems. He stated that the ESBWR derived from earlier GE plant design certification efforts and is the result of eight years of International cooperative work. He stated that the biggest challenge is to cross the regulatory hurdles associated with the inspections, tests, analyses, and acceptance criteria (ITAAC) and combined license (COL) programs. He further stated that he did not know how long it might take to license the ESBWR, in part, because the last GE design certification took about 8 to 10 years. Dr. Rao also provided a brief overview of the GE Nuclear Advance Liquid Metal S-PRISM design

  2. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  3. Standards for safe operation of research reactors

    International Nuclear Information System (INIS)

    1996-01-01

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

  4. Advanced Carbothermal Electric Reactor, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  5. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  6. IEEE No. 323, IEEE trial-use standard: General guide for qualifying Class I electric equipment for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This document describes the basic requirements for the qualification of Class I electric equipment. This is equipment which is essential to the safe shutdown and isolation of the reactor or whose failure or damage could result in significant release of radioactive material. The purpose of this document is to provide guidance for demonstrating the qualifications of electrical equipment as required in the IEEE Std 279 -- Criteria for Nuclear Power Generating Station Protection Systems, and IEEE Std 308 -- Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations. The qualification methods described may be used in conjunction with the Guides for qualifying specific types of equipment, (see Foreword), for updating qualification following modifications or for qualifying equipment for which no applicable Guide exists

  7. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  8. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-07-01

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

  9. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  10. Nuclear reactor development in China for non-electrical applications

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Dong Duo; Xu Yuanhui

    1998-01-01

    In parallel to its vigorous program of nuclear power generation, China has attached great importance to the development of nuclear reactors for non-electrical applications. The Institute of Nuclear Energy Technology (INET) in Beijing has been developing technologies of the water-cooled heating reactor and the modular high temperature gas-cooled reactor. In 1989, a 5 MW water cooled test reactor was erected. Currently, an industrial demonstration nuclear heating plant is being projected. Feasibility studies are being made of sea-water desalination using the INET developed nuclear heating reactor as heat source. Also, a 10 MW high temperature gas-cooled test reactor is being constructed at INET in the framework of China's national high-tech program. The paper gives an overview of China's energy market situation. With respect to China's technology development of high temperature gas-cooled reactors and water cooled heating reactors, the paper describes some general requirements on the technical development, reviews the national programs and activities, describes briefly the design and safety features of the reactor concepts, discusses aspects of application potentials. (author)

  11. Reactor design for nuclear electric propulsion

    International Nuclear Information System (INIS)

    Koenig, D.R.; Ranken, W.A.

    1979-01-01

    Conceptual design studies of a nuclear power plant for electric propulsion of spacecrafts have been on going for several years. An attractive concept which has evolved from these studies and which has been described in previous publications, is a heat-pipe cooled, fast spectrum nuclear reactor that provides 3 MW of thermal energy to out-of-core thermionic converters. The primary motivation for using heat pipes is to provide redundancy in the core cooling system that is not available in gas or liquid-metal cooled reactors. Detailed investigation of the consequences of heat pipe failures has resulted in modifications to the basic reactor design and has led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO 2 and molybdenum sheets that span the entire diameter of the core. Design characteristics are presented and compared for the two reactors

  12. Fire safety requirements for electrical cables towards nuclear reactor safety

    International Nuclear Information System (INIS)

    Raju, M.R.

    2002-01-01

    Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

  13. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  14. Nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-06-01

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

  15. 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

  16. Modular helium reactor for non-electric applications

    International Nuclear Information System (INIS)

    Shenoy, A.

    1997-01-01

    The high temperature gas-cooled Modular Helium Reactor (MHR) is an advanced, high efficiency reactor system which can play a vital role in meeting the future energy needs of the world by contributing not only to the generation of electric power, but also the non-electric energy traditionally served by fossil fuels. This paper summarizes work done over 20 years, by several people at General Atomics, how the Modular Helium Reactor can be integrated to provide different non-electric applications during Process Steam/Cogeneration for industrial application, Process Heat for transportation fuel development and Hydrogen Production for various energy applications. The MHR integrates favorably into present petrochemical and primary metal process industries, heavy oil recovery, and future shale oil recovery and synfuel processes. The technical fit of the Process Steam/Cogeneration Modular Helium Reactor (PS/C-MHR) into these processes is excellent, since it can supply the required quantity and high quality of steam without fossil superheating. 12 refs, 25 figs, 2 tabs

  17. Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Truong Sinh

    1999-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)

  18. Non-electric Applications of Fast Reactors

    International Nuclear Information System (INIS)

    Safa, Henri; Borgard, Jean-Marc

    2013-01-01

    Conclusions: → Most of industrial applications (80%) require low temperature heat below 540°C; → Fast Reactors are technically suitable to provide industrial steam at temperatures not accessible by standard LWRs; → As an illustrative example, the application at an oil refinery site has been studied showing the economic benefits; → Nuclear Cogeneration enhances the overall energy efficiency of the power plant; • Nuclear Cogeneration allows massive cut in CO 2 emissions

  19. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1999-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  20. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  1. General philosophy of safety standards

    International Nuclear Information System (INIS)

    Dunster, H.J.

    1987-01-01

    Safety standards should be related to the form and magnitude of the risk they aim to limit. Because of the lack of direct information at the exposure levels experienced, radiation protection standards have to be based on risk assumptions that, while plausible, are not proven. The pressure for standards has come as much from public perceptions and fears as from the reality of the risk. (author)

  2. Standardized Curriculum for Electricity/Electronics.

    Science.gov (United States)

    Mississippi State Dept. of Education, Jackson. Office of Vocational, Technical and Adult Education.

    Standardized vocational education course titles and core contents are provided for two courses in Mississippi: electricity/electronics I and II. The first course contains the following units: (1) orientation, safety, and leadership; (2) basic principles of electricity/electronics; (3) direct current (DC) theory; (4) magnetism and DC motors; (5)…

  3. 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)

  4. New conducted electrical weapons: Electrical safety relative to relevant standards.

    Science.gov (United States)

    Panescu, Dorin; Nerheim, Max; Kroll, Mark W; Brave, Michael A

    2017-07-01

    We have previously published about TASER ® conducted electrical weapons (CEW) compliance with international standards. CEWs deliver electrical pulses that can inhibit a person's neuromuscular control or temporarily incapacitate. An eXperimental Rotating-Field (XRF) waveform CEW and the X2 CEW are new 2-shot electrical weapon models designed to target a precise amount of delivered charge per pulse. They both can deploy 1 or 2 dart pairs, delivered by 2 separate cartridges. Additionally, the XRF controls delivery of incapacitating pulses over 4 field vectors, in a rotating sequence. As in our previous study, we were motivated by the need to understand the cardiac safety profile of these new CEWs. The goal of this paper is to analyze the nominal electrical outputs of TASER XRF and X2 CEWs in reference to provisions of all relevant international standards that specify safety requirements for electrical medical devices and electrical fences. Although these standards do not specifically mention CEWs, they are the closest electrical safety standards and hence give very relevant guidance. The outputs of several TASER XRF and X2 CEWs were measured under normal operating conditions. The measurements were compared against manufacturer specifications. CEWs electrical output parameters were reviewed against relevant safety requirements of UL 69, IEC 60335-2-76 Ed 2.1, IEC 60479-1, IEC 60479-2, AS/NZS 60479.1, AS/NZS 60479.2, IEC 60601-1 and BS EN 60601-1. Our study confirmed that the nominal electrical outputs of TASER XRF and X2 CEWs lie within safety bounds specified by relevant standards.

  5. The fast reactor and electricity supply, a utility view

    International Nuclear Information System (INIS)

    Wright, J.K.; Hall, R.S.; Kemmish, W.B.; Thorne, R.T.

    1982-01-01

    The significance of the fast reactor is discussed from the viewpoint of the Central Electricity Generating Board. The need for the fast reactor and a possible timescale for its introduction are examined. It is emphasised that demonstration of the commercial and environmental acceptability of the fuel cycle will be needed before any commitment can be made to fast reactors. (U.K.)

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

    International Nuclear Information System (INIS)

    Rico, N.

    1999-01-01

    Full text: The main point of this work consists of a selective method for the engineering of a research reactor based on parameters, which determine a safer design, installation, operation and maintenance. The variety of tasks in a research reactor are: research, development, production of radioisotopes, etc. They are developed within the installation and the different specialties gathered for these activities. It is necessary to count on an intrinsically safe environment, from the point of view of the investigator, the operator and the maintenance personnel. In general, in both nuclear and conventional installation, independent of its size, certain investment necessities prevail, starting from its design, such as: Nuclear Security, Engineering, Versatility, Production (both for investigation and development)), Conventional Security and Physical Protection, Profitability, etc. The concepts which help us accentuate a greater benefit for the research are not found within these parameters, purpose for which this facility was created. When obtaining a simple engineering the results show an increase in security, decrease in maintenance and operative costs, less ageing and an easy operation. The plant engineering of research reactors could be titled, from the engineering and maintenance point of view, as a technological chaos. Not only for its aspect but for its physiognomy too: inaccessible to certain areas; impassable in its circulation aisles; hard to check and measure; disassemble; clean its components; thus increasing unnecessarily the personnel's exposure time. The facilities of research reactors have different disciplines used as rules for the development of the design, such as nuclear, mechanical, thermodynamical, electronic, chemical, electrical, etc. Common guide lines - from design to operation - are non-existent. This is why different manufacturers and models are found within instruments, pumps, electrical engines, illumination, etc. even when they perform

  7. Electrical, instrumentation, and control codes and standards

    International Nuclear Information System (INIS)

    Kranning, A.N.

    1978-01-01

    During recent years numerous documents in the form of codes and standards have been developed and published to provide design, fabrication and construction rules and criteria applicable to instrumentation, control and power distribution facilities for nuclear power plants. The contents of this LTR were prepared by NUS Corporation under Subcontract K5108 and provide a consolidated index and listing of the documents selected for their application to procurement of materials and design of modifications and new construction at the LOFT facility. These codes and standards should be applied together with the National Electrical Code, the ID Engineering Standards and LOFT Specifications to all LOFT instrument and electrical design activities

  8. 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)

  9. Economic aspects of electricity and industrial heat generating reactors

    International Nuclear Information System (INIS)

    Gaussens, J.; Moulle, N.; Dutheil, F.

    1964-01-01

    The economic advantage of electricity-generating nuclear stations decreases when their size decreases. However, when a counter-pressure turbine is joined on to a reactor and the residual heat can be properly used, it can be shown that fairly low capacity nuclear equipment may compete with conventional equipment under certain realistic enough conditions. The aim of this paper is to define these special conditions under which nuclear energy can be profitable. They are connected with the location and the general economic environment of the station, the pattern of the electricity and heat demands it must meet, the level of fuel and specific capital costs, nuclear and conventional. These conditions entail certain technical and economic specifications for the reactors used in this way otherwise they are unlikely to be competitive. In addition, these results are referred to the potential steam and electricity market, which leads us to examine certain uses for the heat generated by double purpose power stations; for example, to supply combined industrial plants, various types of town heating and for removal of salt from sea water. (authors) [fr

  10. Proposal of space reactor for nuclear electric propulsion system

    International Nuclear Information System (INIS)

    Nishiyama, Takaaki; Nagata, Hidetaka; Nakashima, Hideki

    2009-01-01

    A nuclear reactor installed in spacecrafts is considered here. The nuclear reactor could stably provide an enough amount of electric power in deep space missions. Most of the nuclear reactors that have been developed up to now in the United States and the former Soviet Union have used uranium with 90% enrichment of 235 U as a fuel. On the other hand, in Japan, because the uranium that can be used is enriched to below 20%, the miniaturization of the reactor core is difficult. A Light-water nuclear reactor is an exception that could make the reactor core small. Then, the reactor core composition and characteristic are evaluated for the cases with the enrichment of the uranium fuel as 20%. We take up here Graphite reactor, Light-water reactor, and Sodium-cooled one. (author)

  11. 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

  12. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

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

    International Nuclear Information System (INIS)

    Sliter, G.E.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Cokinos, D.M.

    1992-07-01

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

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

    International Nuclear Information System (INIS)

    1981-06-01

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

  16. Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

    International Nuclear Information System (INIS)

    England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

    1990-05-01

    Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

  17. A General Overview of Electric Road Vehicles

    International Nuclear Information System (INIS)

    Lamblin, Veronique

    2018-01-01

    In July 2017 Nicolas Hulot, the French Minister of Ecological and Inclusive Transition, presented a climate plan featuring an end to electricity generation from coal by 2022, a reduction in the nuclear component of electricity supply by one third, a total ban on the sale of petrol or diesel cars by 2040 and an incentive scheme designed gradually to remove polluting vehicles from the roads. Other European partners are following suit and promoting the spread of electric vehicles (Norway, Germany, Netherlands etc.). Yet is this the panacea that will meet the targets for greenhouse gas reduction in the battle against climate change? Futuribles examines the question in this issue with two articles: the first of these by Pierre Bonnaure, above, assesses the forces driving the spread of electric cars and the impediments to that process; this second article by Veronique Lamblin offers a general over - view of electric road vehicles (passenger cars, heavy good vehicles, bicycles etc.) throughout the world. (author)

  18. Thermionic reactor power conditioner design for nuclear electric propulsion.

    Science.gov (United States)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  19. Advanced reactor development for non-electric applications

    International Nuclear Information System (INIS)

    Chang, M.H.; Kim, S.H.

    1996-01-01

    Advance in the nuclear reactor technology achieved through nuclear power programs carried out in the world has led nuclear communities to direct its attention to a better and peaceful utilization of nuclear energy in addition to that for power generation. The efforts for non-electric application of nuclear energy has been pursued in a limited number of countries in the world for their special needs. However, those needs and the associated efforts contributed largely to the development and practical realization of advanced reactors characterized by highly improved reactor safety and reliability by deploying the most up-to-date safety technologies. Due mainly to the special purpose of utilization, economic reasons and ease in implementation of new advanced technologies, small and medium reactors have become a major stream in the reactor developments for non-electric applications. The purpose of this paper is to provide, to the interested nuclear society, the overview of the development status and design characteristics of selected advanced nuclear reactors previously developed and/or currently under development specially for non-electric applications. Major design technologies employed in those reactors to enhance the reactor safety and reliability are reviewed to present the underlying principles of the design. Along with the overview, this paper also introduces a development program and major design characteristics of an advanced integral reactor (SMART) for co-generation purpose currently under conceptual development in Korea. (author)

  20. 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

  1. 5 CFR 2635.202 - General standards.

    Science.gov (United States)

    2010-01-01

    ... 5 Administrative Personnel 3 2010-01-01 2010-01-01 false General standards. 2635.202 Section 2635.202 Administrative Personnel OFFICE OF GOVERNMENT ETHICS GOVERNMENT ETHICS STANDARDS OF ETHICAL... contrary to applicable regulations, policies or guidance relating to the procurement of supplies and...

  2. General remarks on fast neutron reactor physics

    International Nuclear Information System (INIS)

    Barre, J.Y.

    1980-01-01

    The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

  3. General aspects of CAREM-25 reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

    2000-01-01

    CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

  4. General Aspects of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

    2000-01-01

    CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

  5. Reactor power control method upon accidents of electrical power system

    International Nuclear Information System (INIS)

    Hirose, Masao.

    1983-01-01

    Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

  6. Preliminary conceptual design for electrical and I and C system of a new research reactor

    International Nuclear Information System (INIS)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S.

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements

  7. The ISO9000 standard in electrical research

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez-Ortiz, G.; Davila-Nunez, G.; Gonzalez-Sustaeta, J. [Inst. de Investigaciones Electricas, Cuernavaca (Mexico). Unidad de Sistemas Informaticos

    1999-11-01

    In this article, the main aspects that are considered important upon applying the ISO9000 standard [ISO 94] in a research institution, company, or group, are described. The 20 sections of the standard are discussed emphasizing the concepts and definitions that can be differently interpreted in the process of elaboration of a quality manual for a group devoted to research, than when they are applied in the elaboration of a manual for a factory that produces material goods. The exposed reasoning was used in the elaboration of the quality manual in the Information Systems Unit of the Electrical Research Institute in Mexico, and it is being used in the installation and operation of the quality system.

  8. Artificial Intelligence Research at General Electric

    OpenAIRE

    Sweet, Larry

    1985-01-01

    General Electric is engaged in a broad range of research and development activities in artificial intelligence, with the dual objectives of improving the productivity of its internal operations and of enhancing future products and services in its aerospace, industrial, aircraft engine, commercial, and service sectors. Many of the applications projected for AI within GE will require significant advances in the state of the art in advanced inference, formal logic, and architectures for real-tim...

  9. Defining a standard metric for electricity savings

    International Nuclear Information System (INIS)

    Koomey, Jonathan; Akbari, Hashem; Blumstein, Carl; Brown, Marilyn; Brown, Richard; Calwell, Chris; Carter, Sheryl; Cavanagh, Ralph; Chang, Audrey; Claridge, David; Craig, Paul; Diamond, Rick; Eto, Joseph H; Fulkerson, William; Gadgil, Ashok; Geller, Howard; Goldemberg, Jose; Goldman, Chuck; Goldstein, David B; Greenberg, Steve

    2010-01-01

    The growing investment by governments and electric utilities in energy efficiency programs highlights the need for simple tools to help assess and explain the size of the potential resource. One technique that is commonly used in this effort is to characterize electricity savings in terms of avoided power plants, because it is easier for people to visualize a power plant than it is to understand an abstraction such as billions of kilowatt-hours. Unfortunately, there is no standardization around the characteristics of such power plants. In this letter we define parameters for a standard avoided power plant that have physical meaning and intuitive plausibility, for use in back-of-the-envelope calculations. For the prototypical plant this article settles on a 500 MW existing coal plant operating at a 70% capacity factor with 7% T and D losses. Displacing such a plant for one year would save 3 billion kWh/year at the meter and reduce emissions by 3 million metric tons of CO 2 per year. The proposed name for this metric is the Rosenfeld, in keeping with the tradition among scientists of naming units in honor of the person most responsible for the discovery and widespread adoption of the underlying scientific principle in question-Dr Arthur H Rosenfeld.

  10. Defining a standard metric for electricity savings

    Energy Technology Data Exchange (ETDEWEB)

    Koomey, Jonathan [Lawrence Berkeley National Laboratory and Stanford University, PO Box 20313, Oakland, CA 94620-0313 (United States); Akbari, Hashem; Blumstein, Carl; Brown, Marilyn; Brown, Richard; Calwell, Chris; Carter, Sheryl; Cavanagh, Ralph; Chang, Audrey; Claridge, David; Craig, Paul; Diamond, Rick; Eto, Joseph H; Fulkerson, William; Gadgil, Ashok; Geller, Howard; Goldemberg, Jose; Goldman, Chuck; Goldstein, David B; Greenberg, Steve, E-mail: JGKoomey@stanford.ed

    2010-01-15

    The growing investment by governments and electric utilities in energy efficiency programs highlights the need for simple tools to help assess and explain the size of the potential resource. One technique that is commonly used in this effort is to characterize electricity savings in terms of avoided power plants, because it is easier for people to visualize a power plant than it is to understand an abstraction such as billions of kilowatt-hours. Unfortunately, there is no standardization around the characteristics of such power plants. In this letter we define parameters for a standard avoided power plant that have physical meaning and intuitive plausibility, for use in back-of-the-envelope calculations. For the prototypical plant this article settles on a 500 MW existing coal plant operating at a 70% capacity factor with 7% T and D losses. Displacing such a plant for one year would save 3 billion kWh/year at the meter and reduce emissions by 3 million metric tons of CO{sub 2} per year. The proposed name for this metric is the Rosenfeld, in keeping with the tradition among scientists of naming units in honor of the person most responsible for the discovery and widespread adoption of the underlying scientific principle in question-Dr Arthur H Rosenfeld.

  11. Defining a Standard Metric for Electricity Savings

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Marilyn; Akbari, Hashem; Blumstein, Carl; Koomey, Jonathan; Brown, Richard; Calwell, Chris; Carter, Sheryl; Cavanagh, Ralph; Chang, Audrey; Claridge, David; Craig, Paul; Diamond, Rick; Eto, Joseph H.; Fulkerson, William; Gadgil, Ashok; Geller, Howard; Goldemberg, Jose; Goldman, Chuck; Goldstein, David B.; Greenberg, Steve; Hafemeister, David; Harris, Jeff; Harvey, Hal; Heitz, Eric; Hirst, Eric; Hummel, Holmes; Kammen, Dan; Kelly, Henry; Laitner, Skip; Levine, Mark; Lovins, Amory; Masters, Gil; McMahon, James E.; Meier, Alan; Messenger, Michael; Millhone, John; Mills, Evan; Nadel, Steve; Nordman, Bruce; Price, Lynn; Romm, Joe; Ross, Marc; Rufo, Michael; Sathaye, Jayant; Schipper, Lee; Schneider, Stephen H; Sweeney, James L; Verdict, Malcolm; Vorsatz, Diana; Wang, Devra; Weinberg, Carl; Wilk, Richard; Wilson, John; Worrell, Ernst

    2009-03-01

    The growing investment by governments and electric utilities in energy efficiency programs highlights the need for simple tools to help assess and explain the size of the potential resource. One technique that is commonly used in this effort is to characterize electricity savings in terms of avoided power plants, because it is easier for people to visualize a power plant than it is to understand an abstraction such as billions of kilowatt-hours. Unfortunately, there is no standardization around the characteristics of such power plants. In this letter we define parameters for a standard avoided power plant that have physical meaning and intuitive plausibility, for use in back-of-the-envelope calculations. For the prototypical plant this article settles on a 500 MW existing coal plant operating at a 70percent capacity factor with 7percent T&D losses. Displacing such a plant for one year would save 3 billion kW h per year at the meter and reduce emissions by 3 million metric tons of CO2 per year. The proposed name for this metric is the Rosenfeld, in keeping with the tradition among scientists of naming units in honor of the person most responsible for the discovery and widespread adoption of the underlying scientific principle in question--Dr. Arthur H. Rosenfeld.

  12. Status report on nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-01-01

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

  13. IEEE Std 383-1974: IEEE standard for type test of Class IE electric cables, field splices, and connections for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard provides direction for establishing type tests which may be used in qualifying Class 1E electric cables, field splices, and other connections for service in nuclear power generating stations. General guidelines for qualifications are given in IEEE Std 323-1974, Standard for Qualifying Class IE Electric Equipment for Nuclear Power Generating Stations. Categories of cables covered are those used for power control and instrumentation services. Though intended primarily to pertain to cable for field installation, this guide may also be used for the qualification of internal wiring of manufactured devices. This guide does not cover cables for service within the reactor vessel

  14. A standard form for generalized CP transformations

    International Nuclear Information System (INIS)

    Ecker, G.; Grimus, W.; Neufeld, H.

    1987-01-01

    The investigation of general CP transformations leads to transformations of the form U → W T UW with unitary matrices U, W. It is shown that a basis for weak eigenstates can always be chosen such that W T UW has a certain real standard form. (Author)

  15. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

  16. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  17. 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...

  18. Electrical cabling system associated at a nuclear reactor

    International Nuclear Information System (INIS)

    Dejeux, P.; Desfontaines, G.

    1988-01-01

    This cabling system for an electrical device in a nuclear reactor comprises at least a first cable issued of the device, a second cable comprising a first portion, a second portion and a third portion joining the second by a multiple quick fitting connector capable to connect at least ten second portions at ten other third portions of the second cable [fr

  19. Electrical insulators for the theta-pinch fusion reactor

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1976-01-01

    The five major applications for electrical insulators in the Reference Theta Pinch Reactor are as follows: (1) first-wall insulator, (2) blanket intersegment insulator, (3) graphite encapsulating insulator, (4) implosion coil insulator, and (5) compression coil insulator. Insulator design proposals and some preliminary test results are given for each application

  20. Setting a standard for electricity pilot studies

    International Nuclear Information System (INIS)

    Davis, Alexander L.; Krishnamurti, Tamar; Fischhoff, Baruch; Bruine de Bruin, Wandi

    2013-01-01

    In-home displays, dynamic pricing, and automated devices aim to reduce residential electricity use—overall and during peak hours. We present a meta-analysis of 32 studies of the impacts of these interventions, conducted in the US or Canada. We find that methodological problems are common in the design of these studies, leading to artificially inflated results relative to what one would expect if the interventions were implemented in the general population. Particular problems include having volunteer participants who may have been especially motivated to reduce their electricity use, letting participants choose their preferred intervention, and having high attrition rates. Using estimates of bias from medical clinical trials as a guide, we recalculate impact estimates to adjust for bias, resulting in values that are often less than half of those reported in the reviewed studies. We estimate that in-home displays were the most effective intervention for reducing overall electricity use (∼4% using reported data; ∼3% after adjusting for bias), while dynamic pricing significantly reduced peak demand (∼11% reported; ∼6% adjusted), especially when used in conjunction with home automation (∼25% reported; ∼14% adjusted). We conclude with recommendations that can improve pilot studies and the soundness of decisions based on their results. -- Highlights: •We conduct a meta-analysis of field studies of in-home displays, dynamic pricing, and automation on overall and peak use. •Studies were assessed and adjusted for risk-of-bias from inadequate experimental design. •Most studies were at high risk-of-bias from multiple sources. •In-home displays provided the best overall reduction in energy use, approximately 3% after adjustment for risk-of-bias. •Even after adjustment, automation approximately doubled the effectiveness of dynamic pricing on peak reduction from 6% to 14%

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. Design requirement for electrical system of an advanced research reactor

    International Nuclear Information System (INIS)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S.

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system

  8. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  9. Implementation of Electrical Simulation Model for IEC Standard Type-3A Generator

    DEFF Research Database (Denmark)

    Subramanian, Chandrasekaran; Casadei, Domenico; Tani, Angelo

    2013-01-01

    This paper describes the implementation of electrical simulation model for IEC 61400-27-1 standard Type-3A generator. A general overview of the different wind electric generators(WEG) types are given and the main focused on Type-3A WEG standard models, namely a model for a variable speed wind tur...

  10. Standard irradiation facilities for use in TRIGA reactors

    International Nuclear Information System (INIS)

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

    1972-01-01

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

  11. Electrical insulator requirements for mirror fusion reactors

    International Nuclear Information System (INIS)

    Condit, R.H.; Van Konynenburg, R.A.

    1977-01-01

    The requirements for mirror fusion electrical insulators are discussed. Insulators will be required at the neutral beam injectors, injector power supplies, direct converters, and superconducting magnets. Insulators placed at the neutral beam injectors will receive the greatest radiation exposure, 10 14 to 10 16 neutrons/m 2 .s and 0.3 to 3 Gy/s (10 5 to 10 6 R/h) of gamma rays, with shielding. Direct converter insulators may receive the highest temperature (up to 1300 0 K), but low voltage holding requirements. Insulators made from organic materials (e.g., plastics) for the magnet coils may be satisfactory. Immediate conductivity increases of all insulators result from gamma irradiation. With an upper limit to gamma flux exposures of 300 Gy/s in a minimally shielded region, the conductivity could reach 10 -6 S/m. Damage from neutron irradiation may not be serious during several years' exposure. Surface changes in ceramics at the neutral beam injector may be serious. The interior of the injector will contain atomic hydrogen, and sputtering may transfer material away from or onto the ceramic insulators. Unknown and potentially damaging interactions between irradiation, electric fields, temperature gradients, cycling of temperature, surface and joint reactions, sputtering, polarization, and electrotransport in the dielectrics are of concern. Materials research to deal with these problems is needed

  12. Space reactor electric systems: system integration studies, Phase 1 report

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-01-01

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

  13. Electric power from near-term fusion reactors

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Deis, G.A.; Miller, L.G.

    1981-01-01

    This paper examines requirements and possbilities of electric power production on near-term fusion reactors using low temperature cycle technology similar to that used in some geothermal power systems. Requirements include the need for a working fluid with suitable thermodynamics properties and which is free of oxygen and hydrogen to facilitate tritium management. Thermal storage will also be required due to the short system thermal time constants on near-time reactors. It is possbile to use the FED shield in a binary power cycle, and results are presented of thermodynamic analyses of this system

  14. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    Schatz, R.A.; Duetsch, K.L.

    1974-01-01

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  15. Space-reactor electric systems: subsystem technology assessment

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-01-01

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified

  16. 78 FR 56594 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-09-13

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company (GE) GE90-76B, -85B, -90B, -94B, -110B1, and - 115B turbofan engines. This AD was...) Applicability This AD applies to General Electric Company (GE): (1) GE90-76B, -85B, -90B, and -94B turbofan...

  17. 78 FR 72552 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-12-03

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company model GEnx-2B67 and GEnx-2B67B turbofan engines. This AD was prompted by the... certain serial number General Electric Company (GE) model GEnx-2B67 and GEnx-2B67B turbofan engines. The...

  18. Outlines of revised regulation standards for experimental research reactors

    International Nuclear Information System (INIS)

    Hohara, Shinya

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

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

  1. 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

  2. Scaling of silent electrical discharge reactors for hazardous organics destruction

    International Nuclear Information System (INIS)

    Coogan, J.J.; Rosocha, L.A.; Brower, M.J.; Kang, M.; Schmidt, C.A.

    1993-01-01

    Silent electrical discharges are used to produce highly reactive free radicals that destroy hazardous compounds entrained in gaseous effluents at ambient gas temperatures and pressures. We have carried out destruction experiments at Los Alamos on a range of volatile organic compounds (VOCs), including trichloroethylene (TCE), carbon tetrachloride, perchloroethylene (PCE), and chlorofluorocarbons (CFCs). We have measured a ''nine-factor'', the amount of energy required to reduce the VOC concentration by a factor of ten. For practical reactor power densities, the ''nine-factor'' can be used to predict the destruction an removal efficiency (DRE) in terms of gas flow rate and the number of reactor modules. This report proposes a modular, stackable architecture for scaling up the reactor throughput

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  4. 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)

  5. Non-standard interaction effects at reactor neutrino experiments

    International Nuclear Information System (INIS)

    Ohlsson, Tommy; Zhang, He

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    Science.gov (United States)

    George, Jeffrey A.

    1991-01-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

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

    International Nuclear Information System (INIS)

    Berthélemy, Michel; Escobar Rangel, Lina

    2015-01-01

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

  9. The direct conversion of heat into electricity in reactors

    International Nuclear Information System (INIS)

    Devin, B.; Bliaux, J.; Lesueur, R.

    1964-01-01

    The direct conversion of heat into electricity by thermionic emission in an atomic reactor has been studied with the triple aim of its utilisation: as an energy source for a space device, at the head of a conventional conversion system in power installations, or finally in association with the thermoelectric conversion in very low power installations. The laboratory experiments were mainly orientated towards the electron extraction of metals and compounds and their behaviour at high temperatures. Converters furnishing up to 50 amps at 0. 4 volts with an efficiency close to 10 p. 100 have been constructed in the laboratory; the emitters were heated by electron bombardment and were composed of tungsten covered with an uranium carbide deposit or molybdenum covered with cesium. The main aspects of the coupling between the converter and the reactor have been covered from the point of view of electronics: the influence of the mismatching of the load on the temperature of the emitter and the influence of thermal flux density on the temperature of the emitter and the stability of the converter. Converters using uranium carbide as the electron emitter have been tested in reactors. Tests have been made under dynamic conditions in order to determine the dynamic characteristics. The load matching curves have been constructed and the overall performances of several cells coupled in such a way as to form a reactor rod have been deduced. This information is fundamental to the design of a control system for a thermionic conversion reactor. The problems associated with the reliability of thermionic converters connected in series in the same reactor rod have been examined theoretically. Finally, the absorption isotherms have been drawn at the ambient temperatures for krypton and xenon on activated carbon with the aim of investigating the escape of fission products in a converter. (author) [fr

  10. 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)

  11. Nuclear reactor power for an electrically powered orbital transfer vehicle

    Science.gov (United States)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  12. Nuclear reactor power for an electrically powered orbital transfer vehicle

    International Nuclear Information System (INIS)

    Jaffe, L.; Beatty, R.; Bhandari, P.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant

  13. Service to the Electric Utility Industry by the Ford Nuclear Reactor, University of Michigan

    International Nuclear Information System (INIS)

    Burn, R.R.; Simpson, P.A.; Cook, G.M.

    1993-01-01

    Since 1977, the staff of the University of Michigan's Ford Nuclear Reactor has been providing irradiation, testing, analytical, and training services to electric utilities and to suppliers of the nuclear electric utility industry. This paper discusses the reactor's irradiation facilities; reactor programs and utilization; materials testing programs; neutron activation analysis activities; and training programs conducted

  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. Electric Energy Consumption of Multi Purpose Reactor GA. Siwabessy During Reactor Operation

    International Nuclear Information System (INIS)

    Koes Indrakoesoema

    2012-01-01

    Electrical power supply of Reactor Center Multi Purpose obtained from PT PLN to 3030 kVA power contracts. Distribution to existing loads in PRSG divided into 3 (three) lines, each of which is supplied through a transformer BHT01, BHT02 and BHT03, each transformer have capacity of 1600 kVA. During reactor operation, only 2 lines that serve loads, each line serve 2 primary pump motor and 2 secondary pump motor. Electrical power for 24 hours for measurement BHT01, the average is 288 kW, for BHT02 is 641 kW and BHT03 is 466 kW. The energy absorbed by each transformer for 24 hours of measurement, for BHT01 is 6.44 MWh, BHT02 absorb 14.8 MWh and BHT03 absorb 10.9 MWh. (author)

  16. General conditions for electric power supply

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    If it is uncertain whether future power bills will be paid fully, it is admissible to take an action claiming a declaration which states that the electricity rate payment boycotter has no right to non-payment nor a right to withhold payment towards the electricity supply utility, and that the electricity supply utility has the right to stop energy supply because of reduced electricity rate payments effected and/or announced, and to denounce the contract without observing any term of notice. If the electricity buyer reduces a power bill to be paid without any legal grounds, the electricity supply utility has the right to stop power supplies and to denounce the power supply contract without observing any term of notice. The freedom of thought and the freedom of opinion must not be expressed by reducing power bills to be paid. Basic rights discontinue to be effective as soon as a contract or law is broken. A weighing of protected interests is not effected if the exercise of a basic law is unlawful. (orig./HP) [de

  17. 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

  18. Correlation of electrical reactor cable failure with materials degradation

    Energy Technology Data Exchange (ETDEWEB)

    Stuetzer, O.M.

    1986-03-01

    Complete circuit failure (shortout) of electrical cables typically used in nuclear power plant containments is investigated. Failure modes are correlated with the mechanical deterioration of the elastomeric cable materials. It is found that for normal reactor operation, electrical cables are reliable and safe over very long periods. During high temperature excursions, however, cables pulled across corners under high stress may short out due to conductor creep. Severe cracking will occur in short times during high temperatures (>150/sup 0/C) and in times of the order of years at elevated temperatures (100/sup 0/C to 140/sup 0/C). A theoretical treatment of stress distribution responsible for creep and for cracking by J.E. Reaugh of Science Applications, Inc. is contained in the Appendix. 29 refs., 32 figs.

  19. Correlation of electrical reactor cable failure with materials degradation

    International Nuclear Information System (INIS)

    Stuetzer, O.M.

    1986-03-01

    Complete circuit failure (shortout) of electrical cables typically used in nuclear power plant containments is investigated. Failure modes are correlated with the mechanical deterioration of the elastomeric cable materials. It is found that for normal reactor operation, electrical cables are reliable and safe over very long periods. During high temperature excursions, however, cables pulled across corners under high stress may short out due to conductor creep. Severe cracking will occur in short times during high temperatures (>150 0 C) and in times of the order of years at elevated temperatures (100 0 C to 140 0 C). A theoretical treatment of stress distribution responsible for creep and for cracking by J.E. Reaugh of Science Applications, Inc. is contained in the Appendix. 29 refs., 32 figs

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

    International Nuclear Information System (INIS)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao; Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

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

  1. 40 CFR 265.1083 - Standards: General.

    Science.gov (United States)

    2010-07-01

    ... hazardous waste enters the treatment process, any transfer of the hazardous waste is accomplished through continuous hard-piping or other closed system transfer that does not allow exposure of the waste to the... enclosure by conveyor, vehicles, or other mechanical or electrical equipment; or to direct air flow into the...

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

    International Nuclear Information System (INIS)

    1978-11-01

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

  3. 77 FR 4650 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-01-31

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company (GE) CF6-45 and CF6-50 series turbofan engines with certain low-pressure turbine... Compliance We estimate that this AD will affect 387 CF6-45 and CF6-50 series turbofan engines installed on...

  4. 78 FR 72567 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-12-03

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... General Electric Company (GE) GE90-110B1 and -115B turbofan engines. This AD was prompted by multiple... turbofan engines with variable bypass valve (VBV) actuator fuel supply tube, part number (P/N) 2165M22P01...

  5. 78 FR 76045 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-12-16

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... (AD) for General Electric Company (GE) GE90-110B1 and GE90-115B turbofan engines with certain high... turbofan engines with high pressure compressor (HPC) rotor stage 2-5 spools, part numbers (P/Ns) 351-103...

  6. 78 FR 38195 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-06-26

    ... Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT... all General Electric Company (GE) GE90-110B1 and GE90-115B turbofan engines. This emergency AD was.... owners and operators of these GE90-110B1 and GE90-115B turbofan engines. This action was prompted by...

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

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

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

  8. 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

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

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

  11. Electrical characterization and an equivalent circuit model of a microhollow cathode discharge reactor

    International Nuclear Information System (INIS)

    Taylan, O.; Berberoglu, H.

    2014-01-01

    This paper reports the electrical characterization and an equivalent circuit of a microhollow cathode discharge (MHCD) reactor in the self-pulsing regime. A MHCD reactor was prototyped for air plasma generation, and its current-voltage characteristics were measured experimentally in the self-pulsing regime for applied voltages from 2000 to 3000 V. The reactor was modeled as a capacitor in parallel with a variable resistor. A stray capacitance was also introduced to the circuit model to represent the capacitance of the circuit elements in the experimental setup. The values of the resistor and capacitors were recovered from experimental data, and the proposed circuit model was validated with independent experiments. Experimental data showed that increasing the applied voltage increased the current, self-pulsing frequency and average power consumption of the reactor, while it decreased the peak voltage. The maximum and the minimum voltages obtained using the model were in agreement with the experimental data within 2.5%, whereas the differences between peak current values were less than 1%. At all applied voltages, the equivalent circuit model was able to accurately represent the peak and average power consumption as well as the self-pulsing frequency within the experimental uncertainty. Although the results shown in this paper was for atmospheric air pressures, the proposed equivalent circuit model of the MHCD reactor could be generalized for other gases at different pressures.

  12. General safety orientations of the Jules Horowitz Reactor Project (JHRP)

    International Nuclear Information System (INIS)

    Tremodeux, P.; Fiorini, G.L.

    2000-01-01

    After a brief reminder of the JHR purpose, the document outlines the General Safety related Orientations/Recommendations used for the design and the safety assessment of the facility. As far as the JHR design is new, the safety philosophy adopted for this reactor will be as consistent as possible with that recommended for future (power...) reactors. The general recommendations developed in the paper are: the general nuclear safety approach for the design, operation and analysis with, in particular, the adoption of the Defence In Depth principle; the general safety objectives in terms of radiological consequences; the use of Probabilistic Safety Studies; quality assurance. The 'Defence in Depth' concept using amongst others the 'Barrier' principle remains the basis of the JHR safety. 'Defence In Depth' is applied both to design and operation. Its adequacy is checked during the safety assessment and the paper gives the technical recommendations that should allow the designer to implement this concept into the final design. Built mainly for experimental irradiation the JHR facilities will be handled according to conventional or new operation rules which could put materials under stress and entail handling errors. Specific recommendations are defined to take into account the corresponding peculiarities; they are discussed in the paper. The safety design of the JHR takes into account the experience accumulated through the CEA experimental irradiation programmes, which represents several dozen reactor years; the consultation of CEA reactor facilities operators is ongoing. The corresponding feedback is shortly described. Recommendations related to maintenance and associated operation are indicated as well as those regarding the human factor. Details are given on the JHR safety practical implementation through the CEA/DRN Safety approach. Details of the corresponding Safety Objectives are also discussed. Finally the designer position on the role of probabilistic safety

  13. Innovative microbial fuel cell for electricity production from anaerobic reactors

    DEFF Research Database (Denmark)

    Min, Booki; Angelidaki, Irini

    2008-01-01

    A submersible microbial fuel cell (SMFC) was developed by immersing an anode electrode and a cathode chamber in an anaerobic reactor. Domestic wastewater was used as the medium and the inoculum in the experiments. The SMFC could successfully generate a stable voltage of 0.428 ± 0.003 V with a fixed......, a large portion of voltage drop was caused by the ohmic (electrolyte) resistance of the medium present between two electrodes, although the two electrodes were closely positioned (about 3 cm distance; internal resistance = 35 ± 2 Ω). The open circuit potential (0.393 V vs. a standard hydrogen electrode...

  14. Improvement of standards on functional reliability of electric power systems

    International Nuclear Information System (INIS)

    Barinov, V.A.; Volkov, G.A.; Kalita, V.V.; Kogan, F.L.; Makarov, S.F.; Manevich, A.S.; Mogirev, V.V.; Sin'chugov, F.I.; Skopintsev, V.A.; Khvoshchinskaya, Z.G.

    1993-01-01

    Analysis of the most principal aspects of the existing standards and requirements on assuring safety and stability of electric power systems (EPS) and effective (reliable and economical) power supply of consumers is given. The reliability is determined as ability to accomplish the assigned functions. Basic recommendations on improving the standards regulating the safety and reliability of the NPP functioning are formulated

  15. The Effect of Current-Limiting Reactors on the Tripping of Short Circuits in High-Voltage Electrical Equipment

    International Nuclear Information System (INIS)

    Volkov, M. S.; Gusev, Yu. P.; Monakov, Yu. V.; Cho, Gvan Chun

    2016-01-01

    The insertion of current-limiting reactors into electrical equipment operating at a voltage of 110 and 220 kV produces a change in the parameters of the transient recovery voltages at the contacts of the circuit breakers for disconnecting short circuits, which could be the reason for the increase in the duration of the short circuit, damage to the electrical equipment and losses in the power system. The results of mathematical modeling of the transients, caused by tripping of the short circuit in a reactive electric power transmission line are presented, and data are given on the negative effect of a current-limiting resistor on the rate of increase and peak value of the transient recovery voltages. Methods of ensuring the standard requirements imposed on the parameters of the transient recovery voltages when using current-limiting reactors in the high-voltage electrical equipment of power plants and substations are proposed and analyzed

  16. 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)

  17. General aspects to be considered in a research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Grossi, Pablo Andrade; Tello, Cledola Cassia Oliveira de

    2009-01-01

    There are more than 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN-MG) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 kW and it is being licensed for 250 kW, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented general aspects and contents of a Research Reactors Decommissioning Plan. As the Brazilian regulatory body so far does not have a decommissioning policy established neither a regulatory framework in this issue, individual efforts are being integrated to establish a National Decommissioning Group (matrix structure) to perform the decommissioning planning and activities. The approach used for IPR-R1 is presented as suggestions to develop the national regulatory standards on this issue and applied to Brazilian Research Reactors and other nuclear facilities. (author)

  18. 49 CFR 178.915 - General Large Packaging standards.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false General Large Packaging standards. 178.915 Section... PACKAGINGS Large Packagings Standards § 178.915 General Large Packaging standards. (a) Each Large Packaging.... Large Packagings intended for solid hazardous materials must be sift-proof and water-resistant. (b) All...

  19. 40 CFR 63.102 - General standards.

    Science.gov (United States)

    2010-07-01

    ... extent practical. The general duty to minimize emissions during a period of startup, shutdown, or... procedures (including the startup, shutdown, and malfunction plan required in § 63.6(e)(3)), review of... § 63.100(k) of this subpart. [59 FR 19454, Apr. 22, 1994, as amended at 60 FR 63626, Dec. 12, 1995; 61...

  20. Nuclear reactor vessel surface inspecting technique applying electric resistance probe

    International Nuclear Information System (INIS)

    Yamaguchi, T.; Enami, K.; Yoshioka, M.

    1975-01-01

    A new technique for inspecting the inner surface of the PWR type nuclear reactor vessel by use of an electric resistance probe is introduced, centering on a data processing system. This system is composed of a mini-computer, a system typewriter, an interface unit, a D-A converter and controller, and X-Y recorder and others. Its functions are judging flaws and making flaw detection maps. In order to judge flaws by flaw detection signals, three kinds of flaw judging methods have been developed. In case there is a flaw, its position and depth are calculated and listed on the system typewriter. The flaw detection maps are expressed in four kinds of modes and they are displayed on the X-Y recorder. (auth.)

  1. 42 CFR 493.1469 - Standard: Cytology general supervisor qualifications.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Standard: Cytology general supervisor... Nonwaived Testing Laboratories Performing High Complexity Testing § 493.1469 Standard: Cytology general supervisor qualifications. The cytology general supervisor must be qualified to supervise cytology services...

  2. 78 FR 19983 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-04-03

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... Electric Company (GE) CF34-8C and CF34-8E turbofan engines with certain part numbers (P/N) of operability...-8E6, and CF34-8E6A1 turbofan engines, with an operability bleed valve (OBV) part number (P/N...

  3. 78 FR 50320 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-08-19

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... Electric Company (GE) model GEnx-2B67B turbofan engines with booster anti-ice (BAI) air duct, part number...-2B67 turbofan engine be removed from the Applicability section of this AD. The commenters noted that...

  4. 77 FR 3088 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-01-23

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... Electric Company (GE) CF34-10E series turbofan engines. This AD was prompted by a report of heavy wear... turbofan engines installed on airplanes of U.S. registry. We also estimate that it will take about 8 work...

  5. Standard for supply security. A minimum standard to guarantee the balance between electricity demand and supply for the long term

    International Nuclear Information System (INIS)

    Scheepers, M.J.J.; Van Werven, M.J.N.; Seebregts, A.J.; Poort, J.P.; De Nooij, M.; Baarsma, B.E.

    2004-05-01

    The development and use of a minimum reliability standard in the Dutch electricity market to guarantee an adequate balance between electricity demand and supply in the longer term are discussed. This standard can be based on the duration of a power outage and the related costs for society relative to the costs to prevent the power outage. The reliability standard can be translated in an adequacy standard when the reliability of foreign electricity supply to the Dutch market is taken into account. With a theoretical analysis and an assessment of the use of standards in foreign electricity markets and other sectors this study provides a survey of the use of standards in securing public interests. In electricity markets reliability standards can be used obligatory or only to inform market participants of the adequacy of supply preferred by consumers. If no standard is used, the market should rely on the economic incentives provided by contracts and liability. This study proposes to use a reliability standard for calculating the required generation capacity in an ex-ante market analysis using different future scenarios. On the basis of several market indicators, expected market developments can be monitored. Assessment of the market developments relative to the required generation capacity will give a signal to market participants with respect to the expected adequacy in the longer term (7 to 10 years). The assessment and the resulting signal should help to improve market transparency and assist producers, suppliers and consumers in their decisions towards an effective and efficient response on long-term market developments. Market monitoring results can be used by the government to take specific action, if necessary, to reduce barriers to invest. However, more general policy measures should not be linked to the monitoring results since this could provoke strategic behaviour [nl

  6. Local electric dipole moments: A generalized approach.

    Science.gov (United States)

    Groß, Lynn; Herrmann, Carmen

    2016-09-30

    We present an approach for calculating local electric dipole moments for fragments of molecular or supramolecular systems. This is important for understanding chemical gating and solvent effects in nanoelectronics, atomic force microscopy, and intensities in infrared spectroscopy. Owing to the nonzero partial charge of most fragments, "naively" defined local dipole moments are origin-dependent. Inspired by previous work based on Bader's atoms-in-molecules (AIM) partitioning, we derive a definition of fragment dipole moments which achieves origin-independence by relying on internal reference points. Instead of bond critical points (BCPs) as in existing approaches, we use as few reference points as possible, which are located between the fragment and the remainder(s) of the system and may be chosen based on chemical intuition. This allows our approach to be used with AIM implementations that circumvent the calculation of critical points for reasons of computational efficiency, for cases where no BCPs are found due to large interfragment distances, and with local partitioning schemes other than AIM which do not provide BCPs. It is applicable to both covalently and noncovalently bound systems. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  7. General Electric : Immelt surve all / Jena McGregor

    Index Scriptorium Estoniae

    McGregor, Jena

    2008-01-01

    Ettevõtte General Electric juht Jeffrey R. Immelt võitleb selle nimel, et taastada usaldus firma vastu, kuna laenukriis ja kehvad aktsiahinnad on tekitanud olukorra, kus investorid avaldavad suurt survet ettevõtte juhile. Lisa: GE kindralid

  8. General Electric tõrjus Microsofti liidrikohalt / Kaja Koovit

    Index Scriptorium Estoniae

    Koovit, Kaja, 1968-

    2001-01-01

    General Electric võttis Microsofti ees liidrikoha maailma suurima turuväärtusega ettevõtte edetabelis FT500. Tabelid: Maailma suurima turukapitalisatsiooniga ettevõtete TOP25, Skandinaavia firmade turuväärtuse TOP10

  9. EPR, a GEN 3 Reactor providing a competitive electricity cost

    International Nuclear Information System (INIS)

    Salhi, Othman

    2006-01-01

    Since the very beginning of the development of what was to become the EPR, several European entities were involved. The French and German safety authorities expressed that reinforced safety was compulsory. Additional measures were then included to prevent the occurrence of events likely to damage the core, and reduce the possibility of exposure of operating and maintenance personnel. However, not with standing these safety related features resulting from the requirements of the safety authorities, we will focus today on another group of entities that were key players in EPR development: the Utilities. The Utilities voiced their need for a competitive electricity produced and a competitive nuclear reactor. The tradeoff was then to reach both targets in a unique product: a safer and more competitive NPP. Today, the EPR presents features that enable our clients to compete with the cheapest fossil-based electricity production plants. Increased thermal efficiency is obtained both through a higher steam pressure and through careful optimization of the secondary system thermal cycle

  10. Possible Future Role of Small and Medium Sized Reactors (SMRs) in Countries with Small and Medium Electricity Grids

    International Nuclear Information System (INIS)

    Alujevic, L.

    2016-01-01

    Small and Medium Sized Reactors (SMRs) could have enormous potential as options for enhancing the energy supply security, as well as providing a lower capital investment compared to conventional Nuclear Power Plants (NPPs). The trend in SMR development has been towards design certification of small modular reactors, defined as advanced reactors that produce electric power up to 300 MW(e), designed to be built in factories and shipped to utilities for installation as demand arises. The factory-built small modular reactors aim to reduce lengthy construction times while simultaneously increasing quality, thereby minimizing the costs associated with the current time for construction that spans 5 to 8 years. SMR designs include water-cooled reactors, high temperature gas cooled reactors, as well as liquid metal cooled reactors with fast neutron spectrum. Also, many are designed to be emplaced below ground level, giving a high resistance to terrorist threats. The projected timelines of readiness for deployment of SMRs generally range from the present to 2025 - 2030. Currently there are more than 45 SMR designs under development for different application issues. This paper will try to elaborate the benefits and drawbacks of SMRs, as well as describe a couple of designs. Furthermore, some timelines and cost estimates will be provided, depending on the data currently available. Taking all that into account, the conclusion will try to ascertain the suitability of SMRs for Countries with Small and Medium Electricity Grids, namely Croatia. (author).

  11. 78 FR 47534 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-08-06

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... directive (AD) 2013-14-51 for General Electric Company (GE) GE90-110B1 and GE90-115B turbofan engines with... all known U.S. owners and operators of GE90-110B1 and GE90-115B turbofan engines. AD 2013-14-51...

  12. 78 FR 24671 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-04-26

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA... certain General Electric Company (GE) CF6-80C2 series turbofan engines. That AD currently requires.../B1F/B2F/B4F/B6F/B7F/D1F turbofan engines with any of the following installed: (1) Fuel tube, part...

  13. A Comparative Analysis of Three Proposed Federal Renewable Electricity Standards

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, Patrick [National Renewable Energy Lab. (NREL), Golden, CO (United States); Logan, Jeffrey [National Renewable Energy Lab. (NREL), Golden, CO (United States); Bird, Lori [National Renewable Energy Lab. (NREL), Golden, CO (United States); Short, Walter [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2009-05-01

    This paper analyzes potential impacts of proposed national renewable electricity standard (RES) legislation. An RES is a mandate requiring certain electricity retailers to provide a minimum share of their electricity sales from qualifying renewable power generation. The analysis focuses on draft bills introduced individually by Senator Jeff Bingaman and Representative Edward Markey, and jointly by Representative Henry Waxman and Markey. The analysis uses NREL's Regional Energy Deployment System (ReEDS) model to evaluate the impacts of the proposed RES requirements on the U.S. energy sector in four scenarios.

  14. Comparative Analysis of Three Proposed Federal Renewable Electricity Standards

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, P.; Logan, J.; Bird, L.; Short, W.

    2009-05-01

    This paper analyzes potential impacts of proposed national renewable electricity standard (RES) legislation. An RES is a mandate requiring certain electricity retailers to provide a minimum share of their electricity sales from qualifying renewable power generation. The analysis focuses on draft bills introduced individually by Senator Jeff Bingaman and Representative Edward Markey, and jointly by Representative Henry Waxman and Markey. The analysis uses NREL's Regional Energy Deployment System (ReEDS) model to evaluate the impacts of the proposed RES requirements on the U.S. energy sector in four scenarios.

  15. Composite electric generator equipped with steam generator for heating reactor coolant

    International Nuclear Information System (INIS)

    Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato.

    1997-01-01

    The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-12-01

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

  17. Electrical quantum standards and their role in the SI

    Science.gov (United States)

    Robinson, Ian; Georgakopoulos, Dimitrios

    2012-12-01

    realizations of voltage, resistance and current to test their consistency. Active research into this 'metrological triangle' is underway and, at present, there is no evidence to indicate a discrepancy at any level. However, work is continuing on current sources which utilize a countable flow of electrons (the electric current produced is proportional to ef, f being the operating frequency of the device), but the work has some way to go before the question of consistency can be resolved at levels approaching 1 part in 109. In this special feature, Scherer and Camarota review the state-of-the-art of metrological triangle experiments and Devoille et al report on the status of the metrological triangle experiment at the Laboratoire National de Métrologie et d'Essais (LNE), France. The availability of precise representations of the volt and the ohm based on quantum mechanics has enabled the watt balance, an apparatus which relates electrical and mechanical power, to link the kilogram to the Planck constant. This has paved the way for the proposed redefinition of the kilogram, the last artefact standard in the SI, in terms of a fixed value of the Planck constant. In the past few years a number of papers, e.g. [2, 3], have been published describing the working principles of the watt balance and the characteristics of the existing implementations of the experiment. The measurements of the principal quantities—mass, velocity, gravitational acceleration, resistance and voltage—are reasonably well documented but the ultimate precision of the apparatus depends on a number of techniques that are required to eliminate second-order effects. In this special feature, Robinson provides details of these general alignment techniques with special reference to the NPL Mark II watt balance. Acknowledgments We would like to thank the authors for supporting the special feature with their excellent contributions; the guardians of the quality of a scientific paper, the referees, for their valuable

  18. 42 CFR 493.1471 - Standard: Cytology general supervisor responsibilities.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Standard: Cytology general supervisor... Nonwaived Testing Laboratories Performing High Complexity Testing § 493.1471 Standard: Cytology general supervisor responsibilities. The technical supervisor of cytology may perform the duties of the cytology...

  19. 42 CFR 493.837 - Standard; General immunology.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Standard; General immunology. 493.837 Section 493.837 Public Health CENTERS FOR MEDICARE & MEDICAID SERVICES, DEPARTMENT OF HEALTH AND HUMAN SERVICES... These Tests § 493.837 Standard; General immunology. (a) Failure to attain a score of at least 80 percent...

  20. 47 CFR 51.503 - General pricing standard.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 3 2010-10-01 2010-10-01 false General pricing standard. 51.503 Section 51.503... Pricing of Elements § 51.503 General pricing standard. (a) An incumbent LEC shall offer elements to... commission— (1) Pursuant to the forward-looking economic cost-based pricing methodology set forth in §§ 51...

  1. Review of the general atomic experimental fusion power reactor initial conceptual design

    International Nuclear Information System (INIS)

    Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

    1976-01-01

    The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

  2. Renewable Electricity Standards: Good Practices and Design Considerations

    Energy Technology Data Exchange (ETDEWEB)

    Cox, Sadie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Esterly, Sean [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-01-02

    In widespread use globally, renewable electricity standards (RES) are one of the most widely adopted renewable energy policies and a critical regulatory vehicle to accelerate renewable energy deployment. This policy brief provides an introduction to key RES design elements, lessons from country experience, and support resources to enable more detailed and country-specific RES policy design.

  3. 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...

  4. Reactor/Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  5. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  6. 78 FR 41339 - Electric Reliability Organization Proposal To Retire Requirements in Reliability Standards

    Science.gov (United States)

    2013-07-10

    ...] Electric Reliability Organization Proposal To Retire Requirements in Reliability Standards AGENCY: Federal... Reliability Standards identified by the North American Electric Reliability Corporation (NERC), the Commission-certified Electric Reliability Organization. FOR FURTHER INFORMATION CONTACT: Kevin Ryan (Legal Information...

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

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

    2003-01-01

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

  8. Comparison of Standard and Fast Charging Methods for Electric Vehicles

    Directory of Open Access Journals (Sweden)

    Petr Chlebis

    2014-01-01

    Full Text Available This paper describes a comparison of standard and fast charging methods used in the field of electric vehicles and also comparison of their efficiency in terms of electrical energy consumption. The comparison was performed on three-phase buck converter, which was designed for EV’s fast charging station. The results were obtained by both mathematical and simulation methods. The laboratory model of entire physical application, which will be further used for simulation results verification, is being built in these days.

  9. Power reactors in Member States. 1978 edition

    International Nuclear Information System (INIS)

    1978-01-01

    The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

  10. GASSAR-6 (General Atomic Standard Safety Analysis Report). Volume 1

    International Nuclear Information System (INIS)

    1975-01-01

    A standard nuclear steam system for a 3000 MW(t), 1160 MW(e) high temperature gas-cooled reactor (HTGR) nuclear power station is described. The HTGR operates on a uranium-235/thorium-232 cycle. Spherical fuel pellets are coated with multiple layers of pyrolytic carbon, bonded into rods, and encased in hexagonal graphite fuel elements. The core is nuclear-purity-grade, near isotropic graphite machined in hexagonal blocks, which serves as moderator and the heat transfer medium between the fuel and the coolant. Forced helium is the primary coolant and water is the secondary coolant. A prestressed concrete reactor vessel (PCRV) houses the reactor core, primary coolant system and portions of the secondary coolant system, steam generators and circulators. A continuous internal steel liner in the PCRV acts as the primary coolant boundary and sealing membrane. The control rods contain boron carbide in a graphite matrix sheathed in Incoloy 800 cans. Boron constitutes about 40 percent of the absorber material by volume. The control rod drives provide insertion and withdrawal rates consistent with the required reactivity changes for operational load fluctuations and reactor shutdown. Control rods have shim and trip capability. (U.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  12. Design optimization of general arrangement in Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, S. H.; Jung, D. W.; Choi, Y. B.; Cho, S. J.

    1999-01-01

    In order to optimize the general arrangement(GA) of Korean Next Generation Reactor (KNGR), field opinions in domestic nuclear power plants have been collected, and the bench marking on UCN No.1,2 which were estimated to be the most excellent in view of operability and maintenance has been accomplished. Through this work, design optimization items for GA were reviewed. Major items to be selected for optimization are summarized as follows; 'Expanding the compound building function and the mezzanine floor concept in the auxiliary building', 'Including the diesel generator building to the auxiliary building', 'Change of the equipment removal method in the auxiliary building'. With these GA design optimization, the auxiliary building boundary will be improved as a complete rectangular type. The power block volume except the changing effect to the single containment structure will be reduced to about 10% in comparison with that of in KNGR phase II

  13. Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

    International Nuclear Information System (INIS)

    Moraleda Lopez, A.

    1997-01-01

    The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

  14. Summary - Advanced high-temperature reactor for hydrogen and electricity production

    International Nuclear Information System (INIS)

    Forsberg, Charles W.

    2001-01-01

    Historically, the production of electricity has been assumed to be the primary application of nuclear energy. That may change. The production of hydrogen (H 2 ) may become a significant application. The technology to produce H 2 using nuclear energy imposes different requirements on the reactor, which, in turn, may require development of new types of reactors. Advanced High Temperature reactors can meet the high temperature requirements to achieve this goal. This alternative application of nuclear energy may necessitate changes in the regulatory structure

  15. 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

  16. Non-electric applications of pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Adamov, E.O.; Cherkashov, Yu.M.; Romenkov, A.A.

    1997-01-01

    This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

  17. Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

    International Nuclear Information System (INIS)

    Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

    2010-01-01

    The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

  18. 36 CFR 905.735-201 - General standards of conduct.

    Science.gov (United States)

    2010-07-01

    ... economy; (4) Losing complete independence or impartiality; (5) Making a Government decision outside... CORPORATION STANDARDS OF CONDUCT Conduct and Responsibilities of Employees § 905.735-201 General standards of conduct. (a) All employees shall conduct themselves on the job so as to efficiently discharge the work of...

  19. Review of fast reactor operating experience gained in 1998 in Russia. General trends of future fast reactor development

    International Nuclear Information System (INIS)

    Poplavski, V.M.; Ashurko, Y.M.; Zverev, K.V.; Sarayev, O.M.; Oshkanov, N.N.; Korol'kov, A.S.

    1999-01-01

    Review of the general state of nuclear power in Russia as for 1998 is given in brief in the paper. Results of operation of BR-10, BOR-60 and BN-600 fast reactors are presented as well as of scientific and technological escort of the BN-350 reactor. The paper outlines the current status and prospects of South-Urals and Beloyarskaya power unit projects with the BN-800 reactors. The main planned development trends on fast reactors are described concerning both new projects and R and D works. (author)

  20. 77 FR 48110 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-08-13

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Notice of proposed... certain General Electric Company (GE) CF6-80C2 series turbofan engines. The existing AD requires... 2000-04-14, Amendment 39-11597 (65 FR 10698, February 29, 2000), for all GE CF6-80C2 series turbofan...

  1. 76 FR 64844 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2011-10-19

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Notice of proposed... General Electric Company (GE) CF6-45 and CF6-50 series turbofan engines with certain low-pressure turbine... series turbofan engines with certain LPT rotor stage 3 disks installed. That AD requires initial and...

  2. 77 FR 58471 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-09-21

    ... Airworthiness Directives; General Electric Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA.../P1, GEnx-1B75/P1, GEnx- 2B67, and GEnx-2B67B turbofan engines. This AD requires initial and... this AD will affect 11 GE GEnx turbofan engines installed on airplanes of U.S. registry. We also...

  3. 77 FR 76977 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2012-12-31

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Supplemental notice... proposed airworthiness directive (AD) for certain General Electric Company (GE) CF6-80C2 series turbofan... part 39 to include an AD that would apply to certain GE CF6-80C2 series turbofan engines. That NPRM...

  4. power quality study of the electrical system of the second research reactor of egypt

    International Nuclear Information System (INIS)

    Fathi, H.M.E.

    2007-01-01

    identifying problems in an electrical network which may results in an extensive damage occur is advisable for any power system. monitoring of power quality is necessary for maintaining accurate operation of sensitive equipment, it also ensures that unnecessary for maintaining accurate operation of sensitive equipment, it also ensures that unnecessary energy losses in a power system are kept at a minimum which lead to more profits. with advanced in technology growing of industrial/commercial facilities in many region. power quality problems have been a major concern among engineers; particularly in an industrial environment, where there are many large-scale type of equipment.thus, it would be useful to monitor and analyze the power quality problems.determination of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. the choice of monitoring equipment in a survey is also important to ascertain a solution to these power quality problems. a power quality survey involves gathering data resources; analyzing the data (with reference to power quality standard); then , if problems exist, recommendation of mitigation techniques. the main objective of the present work is to identify power quality problems in second research reactor of egypt (ETRR-2). electrical power is supplied to the installation via tow incoming feeders from tow different substations (incoming feeder from new Abo-Zabal substation and incoming feeder from old Abo-Zabal substation ) to keep good reliability , each feeder is designed to carry the full load of ETRR-2. the monitoring investigations were performed at the second research reactor of egypt for both feeders at different operation conditions

  5. A design study of high electric power for fast reactor cooled by supercritical light water

    International Nuclear Information System (INIS)

    Koshizuka, Seiichi

    2000-03-01

    In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600 MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212 MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about 1.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power. (author)

  6. Education for university students, high school teachers and the general public using the Kinki University Reactor

    International Nuclear Information System (INIS)

    Tsuruta, T.

    2007-01-01

    Atomic Energy Research Institute of Kinki University is equipped with a nuclear reactor which is called UTR-KINKI. UTR is the abbreviation for University Teaching and Research Reactor. The reactor is the first one installed in Japanese universities. Though the reactor is owned and operated by Kinki University, its use is widely open to scientists and students from other universities and research institutions. The reactor is made the best of teaching instrument for the training of high school teachers. In addition, the reactor is utilized for general public education concerning atomic energy. (author)

  7. International standard for the charging of electric vehicles; Internationaler Standard fuer das Laden von Elektrofahrzeugen - Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Mathoy, A.

    2001-12-15

    This final report for the Swiss Federal Office of Energy (SFOE) reports on the situation regarding the standardisation of battery charging systems for electric vehicles. The advantages of an international standard both for users and manufacturers of electrically-powered vehicles are discussed. The work done in the IEC and CENELEC technical committees is reviewed. Developments achieved since 1999 are reviewed and further developments and work to be done are examined. The most important points in the IEC standard 61851 are looked at and various connector interfaces are described. Direct and inductive charging systems according to IEC 61980 are examined and the special situation in Italy, where power available in homes for the charging of electrical vehicles is more limited, is reviewed.

  8. Research and development activities for reactor decommissioning. Developing technology of Fuji Electric Co., Ltd

    International Nuclear Information System (INIS)

    Shirakawa, Masahiro; Takaya, Jyunichi; Mizukoshi, Seiji; Hosoda, Hiroshi; Tomizuka, Chiaki; Funaguchi, Susumu; Ito, Katsuhito

    1997-01-01

    Fuji Electric Co., Ltd. is conducting decommissioning R and D for commercial reactor, especially for gas cooled reactor since the construction of the Tokai-1 power station of JAPCO, in the field of system engineering, residual radioactivity evaluation, dismantling of core internals, remote handling, treatment and disposal of radioactive waste, and radioactivity measurement. These R and D have been performed mainly under contract of JAPCO and JAERI. This paper gives a summary of the present status and future plan concerning technical development for decommissioning of nuclear reactor by Fuji Electric Co., Ltd. (author)

  9. General areas needing chemical competence to support reactor operation

    International Nuclear Information System (INIS)

    Proksch, E.; Bildstein, H.

    1963-01-01

    Chemical competence is needed not only for the development of new types of reactors but also for the start-up and safe operation of reactors. The activities of chemistry and chemical engineering cover a number of fields, namely chemical analysis, radiochemical analysis, corrosion research, radiolysis of water and water purification. The author reviews fields in reactor operation and maintenance in which chemical competence is needed. (author). 9 refs

  10. 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.

  11. Magnox Electric Littlebrook reactor inspection and repair rehearsal facility

    International Nuclear Information System (INIS)

    Barnes, S.A.; Clayton, R.; Gaydon, B.G.; Ramsey, B.H.

    1996-01-01

    Magnox reactors, although designed to be maintenance free during their operational life, have nevertheless highlighted the need for test rig facilities to train operators in the methods and techniques of reactor inspection and repair. The history of the facility for reactor engineering development (FRED) is described and its present role as a repair rehearsal facility noted. Advances in computer graphics may, in future, mean that such operator training will be virtual reality rather than analog reality based; however the need for such rigs to commission techniques and equipment and to establish performance and reliability is likely to continue. (UK)

  12. Nuclear electric capacity expansion in Mexico: system effects of reactor size and cost

    International Nuclear Information System (INIS)

    Thayer, G.R.; Abbey, D.S.; Hardie, R.W.; Enriquez, R.P.; Uria, E.G.

    1984-01-01

    Mexico's electrical generation capacity could more than double over the next ten years - from about 15 GWe currently to as much as 35 GWe in 1990. While new capacity additions will be predominantly oil-fired in the 1980's, nuclear power will become increasingly important in the 1990's. This study investigated the appropriate size of new, nuclear capacity additions by assessing the implications of installing different size reactors into Mexico's electrical grid. Included in the assessments of reactor sizes are estimates of electrical generation costs and comparisons of the effective load-carrying capability of a 10 GWe nuclear capacity expansion

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  14. Small Modular Reactors for Countries with Small to Medium Electric Grids - An Economically Sensible Solution

    International Nuclear Information System (INIS)

    Young, P.

    2012-01-01

    There has recently been a renewed interest throughout the world in small nuclear units for generating electricity and for other applications. A report by the World Nuclear Association discussing the advantages of small modular nuclear reactors (SMRs) over traditional nuclear reactor designs, states that ''modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production, and reduced siting costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction.'' Since the inception of nuclear power, the size of reactor units has grown from under 100 MWe to more than 1600 MWe. Today, due partly to the high capital cost of large power reactors and partly to the need to service small electricity grids, there is a move to develop smaller units. These may be built individually or as modules in a larger plant. SMRs are a good fit in markets where anticipated electricity demand is projected to increase incrementally, because SMRs could be built in series as needed. SMRs might be particularly attractive in countries that currently rely on diesel generators for producing electricity. Small reactors could make economic sense because of the high cost of diesel generation compared to the low marginal cost of producing electricity from nuclear energy. (Keeping in mind the initial investment costs and the need to establish a national regulatory program.) Some SMR designs are fabricated in a factory and then delivered to the site. This could be a solution for markets that lack the qualified engineers and skilled craft workers needed to construct large reactors on site. This paper will provide an overview of the types and attributes of SMRs in use or under development worldwide, describe the similarities and important differences between designs, discuss potential applications for SMRs, including baseload electricity generation, electricity generation for remote locations and areas with

  15. Development of conductor feedthrough module of LV electrical penetration assembly for research reactors

    International Nuclear Information System (INIS)

    Luo Zhiyuan; Wang Guangjin; Zhou Bin

    2007-01-01

    A LV electrical penetration assembly with perfusion sealing conductor feedthrough module was developed, which can be used for the connection of internal and external cables through the wall of the research reactor workshop. The LV electrical penetration assembly was combined with several independent modules. The maintenance and replacement of the assembly can be easily done in service. The sealing of conductor feedthrough module was achieved with the perfusion of self-extinguishing epoxy. The leakage between the conductor feedthrough module and the end plate module was blocked with rubber rings. The result of the leakage test and the electrical performance test for the samples of conductor feedthrough module satisfied the requirement of research reactor. The structure of the new electrical penetration assembly is simple and compact. It can be manufactured with mature technology and cost low price. The performance of the assembly is steady. It can be used widely in research reactors. (authors)

  16. Radwaste volume reduction and solidification by General Electric

    International Nuclear Information System (INIS)

    Green, T.A.; Weech, M.E.; Miller, G.P.; Eberle, J.W.

    1982-01-01

    Since 1978 General Electric has been actively engaged in developing a volume reduction and solidifcation system or treatment of radwaste generated in commercial nuclear power plants. The studies have been aimed at defining an integrated system that would be directly responsive to the rapid evolving needs of the industry for the volume reduction and solidification of low-level radwaste. The resulting General Electric Volume Reduction System (GEVRS) is an integrated system based on two processes: the first uses azeotropic distillation technology and is called AZTECH, and the second is controlled-air incineration...called INCA. The AZTECH process serves to remove water from concentrated salt solutions, ion exchange resins and filter sludge slurries and then encapsulates the dried solids into a dense plastic product. The INCA unit serves to reduce combustible wastes to ashes suitable for encapsulation into the same plastic product produced by AZTECH

  17. 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

  18. Neutron electric dipole moment and extension of the standard model

    International Nuclear Information System (INIS)

    Oshimo, Noriyuki

    2001-01-01

    A nonvanishing value for the electric dipole moment (EDM) of the neutron is a prominent signature for CP violation. The EDM induced by the Kobayashi-Maskawa mechanism of the standard model (SM) has a small magnitude and its detection will be very difficult. However, since baryon asymmetry of the universe cannot be accounted for by the SM, there should exist some other source of CP violation, which may generate a large magnitude for the EDM. One of the most hopeful candidates for physics beyond the SM is the supersymmetric standard model, which contains such sources of CP violation. This model suggests that the EDM has a magnitude not much smaller than the present experimental bounds. Progress in measuring the EDM provides very interesting information about extension of the SM. (author)

  19. Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

    International Nuclear Information System (INIS)

    Forsberg, C.

    2015-01-01

    The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

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

    International Nuclear Information System (INIS)

    Heising-Goodman, C.D.

    1981-01-01

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

  1. Implementation of the rapid cross section adjustment approach at General Electric

    International Nuclear Information System (INIS)

    Cowan, C.L.; Kujawski, E.; Protsik, R.

    1978-01-01

    The General Electric rapid cross section adjustment approach was developed to use the shielding factor method for formulating multigroup cross sections. In this approach, space- and composition-dependent cross sections for a particular reactor or shield design are prepared from a generalized cross section library by the use of resonance self-shielding factors, and by the adjustment of elastic scattering cross sections for the local neutron flux spectra. The principal tool in the cross section adjustment package is the data processing code TDOWN. This code was specified to give the user a high degree of flexibility in the analysis of advanced reactor designs. Of particular interest in the analysis of critical experiments is the ability to carry out cell heterogeneity self-shielding calculations using a multiregion equivalence relationship, and the homogenization of the cross sections over the specified cell with the flux weighting obtained from transport theory calculations. Extensive testing of the rapid cross section adjustment approach, including comparisons with Monte Carlo methods, indicated that this approach can be utilized with a high degree of confidence in the design analysis of complex fast reactor systems. 2 figures, 1 table

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

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

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

  3. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Science.gov (United States)

    2013-09-30

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

  4. Evaluation and optimization of General Atomics' GT-MHR reactor cavity cooling system using an axiomatic design approach

    International Nuclear Information System (INIS)

    Thielman, Jeff; Ge, Ping; Wu, Qiao; Parme, Laurence

    2005-01-01

    The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units

  5. General formulation of standard model the standard model is in need of new concepts

    International Nuclear Information System (INIS)

    Khodjaev, L.Sh.

    2001-01-01

    The phenomenological basis for formulation of the Standard Model has been reviewed. The Standard Model based on the fundamental postulates has been formulated. The concept of the fundamental symmetries has been introduced: To look for not fundamental particles but fundamental symmetries. By searching of more general theory it is natural to search first of all global symmetries and than to learn consequence connected with the localisation of this global symmetries like wise of the standard Model

  6. 75 FR 17036 - Energy Conservation Program: Energy Conservation Standards for Small Electric Motors; Correction

    Science.gov (United States)

    2010-04-05

    ... Conservation Program: Energy Conservation Standards for Small Electric Motors; Correction AGENCY: Office of... standards for small electric motors, which was published on March 9, 2010. In that final rule, the U.S... titled ``Energy Conservation Standards for Small Electric Motors.'' 75 FR 10874. Since the publication of...

  7. 10 CFR 431.446 - Small electric motors energy conservation standards and their effective dates. [Reserved

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 3 2010-01-01 2010-01-01 false Small electric motors energy conservation standards and... CONSERVATION ENERGY EFFICIENCY PROGRAM FOR CERTAIN COMMERCIAL AND INDUSTRIAL EQUIPMENT Small Electric Motors Energy Conservation Standards § 431.446 Small electric motors energy conservation standards and their...

  8. 78 FR 73589 - Energy Conservation Program: Energy Conservation Standards for Commercial and Industrial Electric...

    Science.gov (United States)

    2013-12-06

    ... Conservation Program: Energy Conservation Standards for Commercial and Industrial Electric Motors; Proposed... Conservation Program: Energy Conservation Standards for Commercial and Industrial Electric Motors AGENCY... proposes energy conservation standards for a number of different groups of electric motors that DOE has not...

  9. 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

  10. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

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

  11. 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)

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

    International Nuclear Information System (INIS)

    Ramana, M.V.; Saikawa, Eri

    2011-01-01

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

  13. Wiring installation for electric devices above the roof slab of a nuclear reactor

    International Nuclear Information System (INIS)

    Jahnke, S.

    1986-01-01

    The wiring installation is situated inside the nuclear reactor building. It includes, associated to electric devices, a first cable which extends from the device to a fixed connector arranged above the cover. A second cable is connected to the said fixed connector and to a connector fixed on a plate situated out of the reactor. According to the present invention each second cable has several sections. A first section can be connected to the said fixed connector situated above the cover and to a fixed lead-in connector of a fluid-tight conduit above the reactor core. A second section is inside the conduit. A third section can be connected to a lead-out connector fixed on the plate which is out of the reactor. The invention applies more particularly to pressurized water nuclear reactors [fr

  14. The Integral Fast Reactor concept: Today's hope for tomorrow's electrical energy needs

    International Nuclear Information System (INIS)

    Dwight, C.C.; Phipps, R.D.

    1989-01-01

    Acid rain and the greenhouse effect are getting more attention as their impacts on the environment become evident around the world. Substantial evidence indicates that fossil fuel combustion for electrical energy production activities is a key cause of those problems. A change in electrical energy production policy is essential to a stable, healthy environment. That change is inevitable, it's just a matter of when and at what cost. Vision now, instead of reaction later, both in technological development and public perception, will help to limit the costs of change. The Integral Fast Reactor (IFR) is a visionary concept developed by Argonne National Laboratory that involves electrical energy production through fissioning of heavy metals by fast neutrons in a reactor cooled by liquid sodium. Physical characteristics of the coolant and fuel give the reactor impressive characteristics of inherent and passive safety. Spent fuel is pyrochemically reprocessed and returned to the reactor in the IFR's closed fuel cycle. Advantages in waste management are realized, and the reactor has the potential for breeding, i.e., producing as much or more fuel than it uses. This paper describes the IFR concept and shows how it is today's hope for tomorrow's electrical energy needs. 14 refs., 1 fig., 1 tab

  15. Nuclear reactor capable of electric power generation during in-service inspection

    International Nuclear Information System (INIS)

    Nakamura, Shinsuke; Nogami, Hitoshi.

    1992-01-01

    The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

  16. Neutron electric dipole moment in the minimal supersymmetric standard model

    International Nuclear Information System (INIS)

    Inui, T.; Mimura, Y.; Sakai, N.; Sasaki, T.

    1995-01-01

    The neutron electric dipole moment (EDM) due to the single quark EDM and to the transition EDM is calculated in the minimal supersymmetric standard model. Assuming that the Cabibbo-Kobayashi-Maskawa matrix at the grand unification scale is the only source of CP violation, complex phases are induced in the parameters of soft supersymmetry breaking at low energies. The chargino one-loop diagram is found to give the dominant contribution of the order of 10 -27 similar 10 -29 e.cm for the quark EDM, assuming the light chargino mass and the universal scalar mass to be 50 GeV and 100 GeV, respectively. Therefore the neutron EDM in this class of model is difficult to measure experimentally. The gluino one-loop diagram also contributes due to the flavor changing gluino coupling. The transition EDM is found to give dominant contributions for certain parameter regions. (orig.)

  17. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    Rust, J.H.; Clement, J.D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  18. Use of small reactors as an alternative to supply electricity to Baja California Sur

    International Nuclear Information System (INIS)

    Alonso, G.; Portes, E.; Ramirez, J. R.; Ortega, G.

    2016-09-01

    The state of Baja California Sur (Mexico) does not form part of the national interconnected electrical system of the country, reason why is local its electrical power supply; one of the alternatives to cover future demands is the use of gas-based combined cycles, which presents the additional problem of including a high price for gas transportation in its costs. In order to reduce total costs, including investment, fuels and operation and maintenance in the operation of the Baja California Sur state electricity system in the coming years, mainly due to the estimated natural gas cost order of $11.50 dollars per million BTU, a proposal is presented to reduce the costs of the electrical system by replacing the necessary combined cycles with the new Small Modular Reactor type nuclear reactors, this alternative is economically competitive. (Author)

  19. NEPTUNIX, a general program of simulation applied to nuclear reactors

    International Nuclear Information System (INIS)

    Bonnemay, A.; Dansac Bon, V.

    1978-01-01

    Most simulation languages admit an incremental description and involve explicit integration algorithms. NEPTUNIX is a simulation language directly admitting algebraic differential equations under an implicit form, and it involves a very efficient implicit integration method with variable step and order. NEPTUNIX is a tool used for building large systems models in the field of nuclear reactors [fr

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

    Science.gov (United States)

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

    2018-08-01

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

  1. Power unit with GT-MHR reactor plant for electricity production and district heating

    International Nuclear Information System (INIS)

    Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

    2000-01-01

    Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

  2. Electrical measuring device for a high temperature reactor

    International Nuclear Information System (INIS)

    Elter, C.; Handel, H.; Schoening, J.; Schmitt, H.

    1982-01-01

    The device for measuring the low or high neutron flux during start-up or at load is accommodated in an armoured guide tube projecting into the floor. A gas-tight capsule is formed as the measuring column with outer dome with a lid solidly connected by a flange to the armoured tube situated on the side wall of the concrete reactor vessel, together with the armoured guide tube. Two shielding shutters prevent the passage of radiation through the armoured tube. (DG) [de

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Neutron importance and the generalized Green function for the conventionally critical reactor with normalized neutron distribution

    International Nuclear Information System (INIS)

    Khromov, V.V.

    1978-01-01

    The notion of neutron importance when applied to nuclear reactor statics problems described by time-independent homogeneous equations of neutron transport with provision for normalization of neutron distribution is considered. An equation has been obtained for the function of neutron importance in a conditionally critical reactor with respect to an arbitrary nons linear functional determined for the normalized neutron distribution. Relation between this function and the generalized Green function of the selfconjugated operator of the reactor equation is determined and the formula of small perturbations for the functionals of a conditionally critical reactor is deduced

  5. General scheme of research reactor mainly for production of fission 99Mo

    International Nuclear Information System (INIS)

    Shen Feng; Liu Xingmin; Wu Xiaochun; Sun Zheng; Guo Chunqiu; Yi Dayong

    2012-01-01

    On the basis of the analysis for current circumstance and development tendency of research reactor mainly for 99 Mo production in the world, the design idea of this sort of research reactor was proposed. By the optimization and basic design, the general design parameters of the reactor were analyzed and testified. The evaluation of output activities of 99 Mo and the analysis of economics were conducted on the basically assumption. It is argued that the economics of this reactor is improved dramatically while the safety is ensured by the analysis. (authors)

  6. Conceptual design of a demonstration reactor for electric power generation

    International Nuclear Information System (INIS)

    Asaoka, Y.; Hiwatari, R.; Okano, K.; Ogawa, Y.; Ise, H.; Nomoto, Y.; Kuroda, T.; Mori, S.; Shinya, K.

    2005-01-01

    Conceptual study on a demonstration plant for electric power generation, named Demo-CREST, was conducted based on the consideration that a demo-plant should have capacities both (1) to demonstrate electric power generation in a plant scale with moderate plasma performance, which will be achieved in the early stage of the ITER operation, and foreseeable technologies and materials and (2) to have a possibility to show an economical competitiveness with advanced plasma performance and high performance blanket systems. The plasma core was optimized to be a minimum size for both net electric power generation with the ITER basic plasma parameters and commercial-scale generation with advance plasma parameters, which would be attained by the end of ITER operation. The engineering concept, especially the breeding blanket structure and its maintenance scheme, is also optimized to demonstrate the tritium self-sustainability and maintainability of in-vessel components. Within the plasma performance as planned in the present ITER program, the net electric power from 0 MW to 500 MW is possible with the basic blanket system under the engineering conditions of maximum magnetic field 16 T, NBI system efficiency 50%, and NBI current drive power restricted to 200 MW. Capacities of stabilization of reversed shear plasma and the high thermal efficiency are additional factors for optimization of the advanced blanket. By replacing the blanket system with the advanced one of higher thermal efficiency, the net electric power of about 1000 MW is also possible so that the economic performance toward the commercial plant can be also examined with Demo-CREST. (author)

  7. Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

    International Nuclear Information System (INIS)

    Mitari, O.; Hirose, A.; Skarsgard, H.M.

    1989-01-01

    In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

  8. Pump selection and application in a pressurized water reactor electric generating plant

    International Nuclear Information System (INIS)

    Kitch, D.M.

    1985-01-01

    Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

  9. A European view of the use of nuclear reactors for applications other than electricity generation

    International Nuclear Information System (INIS)

    Marsham, T.N.; Brierley, G.

    Energy demands and temperature ranges employed by heat-consuming industrial processes are analyzed. Matching heat demand to reactor size is a problem. Emphasis is placed on HTGR's providing heat in the range 300-800 deg C. Further non-electrical uses of nuclear power, like nuclear ship propulsion, are analyzed. (E.C.B.)

  10. An electrical pulse hydride injector (EPHI) for reactor fueling and tritium handling applications

    International Nuclear Information System (INIS)

    Azizov, E.A.; Kareev, Yu.A.; Savotkin, A.N.; Frunze, V.V.; Penzhorn, R.D.; Glugla, M.

    1995-01-01

    An electrical pulse hydride injector (EPHI) has been developed for reactor fuelling as well as for handling of hydrogen isotopes in facilities operating with tritium. Salient features of the EPHI are the accuracy with which the fuelling rate can be controlled and the avoidance of a pressurized ballast. The generator is simple and allows for safe operation with tritium. (orig.)

  11. General considerations in fuel management for thermal reactors

    International Nuclear Information System (INIS)

    Tyror, J.G.; Fayers, F.J.

    1971-07-01

    By fuel management we mean the strategy for fuelling and refuelling a reactor together with any associated absorber movements. It incorporates (a) decisions made about the timing of fuel loading operations; (b) choice of enrichments to be loaded; (c) selection of sites at which reloading occurs; (d) programming of control rods and any other reactivity control facilities such as soluble or burnable poisons; and (e) evaluation of the resulting fuel element performance consequences. The topic of fuel management is thus a vast and vital one. It embraces most of the various aspects of core performance and determines many of a reactor's design characteristics. In this paper we review what to us appear to be some of the important issues in this important field

  12. General considerations for neutron capture therapy at a reactor facility

    International Nuclear Information System (INIS)

    Binney, S.E.

    2001-01-01

    In addition to neutron beam intensity and quality, there are also a number of other significant criteria related to a nuclear reactor that contribute to a successful neutron capture therapy (NCT) facility. These criteria are classified into four main categories: Nuclear design factors, facility management and operations factors, facility resources, and non-technical factors. Important factors to consider are given for each of these categories. In addition to an adequate neutron beam intensity and quality, key requirements for a successful neutron capture therapy facility include necessary finances to construct or convert a facility for NCT, a capable medical staff to perform the NCT, and the administrative support for the facility. The absence of any one of these four factors seriously jeopardizes the overall probability of success of the facility. Thus nuclear reactor facility management considering becoming involved in neutron capture therapy, should it be proven clinically successful, should take all these factors into consideration. (author)

  13. High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2016-01-01

    Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  14. Generation of net electric power with a tokamak reactor under foreseeable physical and engineering conditions

    International Nuclear Information System (INIS)

    Hiwatari, R.; Asaoka, Y.; Okano, K.; Yoshida, T.; Tomabechi, K.

    2004-01-01

    This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with β N = 1.8, HH = 1.0, andf nGW 0.85 had sufficient potential to achieve the electric break-even condition (net electric power P e net = 0MW) under the following engineering conditions: machine major radius 6.5m ≤ R p ≤ 8.5m, the maximum magnetic field on TF coils B tmax = 16 T, thermal efficiency η e 30%, and NBI system efficiency η NBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are β N and fη GW . ≥ 3.0 is required for P e net ∼ 600MW with fusion power P f ∼ 3000MW. On the other hand, fη GW ≥ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of T ave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs P f (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius R p ∼ 7.5m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy. (author)

  15. 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.

  16. The fast breeder reactor and the electricity provision system

    International Nuclear Information System (INIS)

    1975-01-01

    Due to inflexible planning, the electricity provision system is bound to generate demands for big centralized power plants, that is, nuclear power plants. A lack of feed-back from producer to consumer results in a technical distribution net-work that cannot react to big changes in supply and demand. This makes it particularly difficult to include alternative energy sources with fluctuating energy production

  17. Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

    International Nuclear Information System (INIS)

    PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

    2003-01-01

    The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

  18. Selection of power plant elements for future reactor space electric power systems

    International Nuclear Information System (INIS)

    Buden, D.; Bennett, G.A.; Copper, K.

    1979-09-01

    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

  19. The role of the Small Modular Reactors for isolated electric systems in Brazil

    International Nuclear Information System (INIS)

    Mattos, João Roberto Loureiro de; Dias, Marcio Soares

    2017-01-01

    Small modular reactors (SMRs) are being considered as a useful option to reduce the total capital costs of nuclear power plants nad to provide power in small or off-grid systems. The SMRs offer simpler, standardized and safer modular design by being factory built, requiring smaller initial capital investment, and having shorter construction times. SMRs could be small enough to be transportable, could be used in isolated locations without advanced infrastructure and without power grid, or could be clustered in a single site to provide a multi-module, large capacity power plant. In this sense, Brazil has about 200 isolated systems in the Amazon region with no connection to the national grid. SMRs could be suitable to serve some of these smaller energy markets with less developed infrastructure, replacing large diesel generators, with expensive fuel cost, often noisy, highly polluting and have low overall efficiency. Despite the economics of SMRs have yet to be proven, the Brazilian growth of electricity generation will demand all available sources of energy and some propose not to discard their evaluation in terms of SMR applications. This paper addresses the potential application of SMRs with focus in specific Brazilian regions. (author)

  20. The role of the Small Modular Reactors for isolated electric systems in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Mattos, João Roberto Loureiro de; Dias, Marcio Soares, E-mail: jrmattos@cdtn.br, E-mail: marciod@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Small modular reactors (SMRs) are being considered as a useful option to reduce the total capital costs of nuclear power plants nad to provide power in small or off-grid systems. The SMRs offer simpler, standardized and safer modular design by being factory built, requiring smaller initial capital investment, and having shorter construction times. SMRs could be small enough to be transportable, could be used in isolated locations without advanced infrastructure and without power grid, or could be clustered in a single site to provide a multi-module, large capacity power plant. In this sense, Brazil has about 200 isolated systems in the Amazon region with no connection to the national grid. SMRs could be suitable to serve some of these smaller energy markets with less developed infrastructure, replacing large diesel generators, with expensive fuel cost, often noisy, highly polluting and have low overall efficiency. Despite the economics of SMRs have yet to be proven, the Brazilian growth of electricity generation will demand all available sources of energy and some propose not to discard their evaluation in terms of SMR applications. This paper addresses the potential application of SMRs with focus in specific Brazilian regions. (author)

  1. Controlling Electrical Hazards

    National Research Council Canada - National Science Library

    1997-01-01

    ...). In general, OSHA's electrical standards are based on the National Fire Protection Associations Standard NFPA 70E, Electrical Safety Requirements for Employee Workplaces, and in turn, from the National Electrical Code (NEC...

  2. 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

  3. 76 FR 58101 - Electric Reliability Organization Interpretation of Transmission Operations Reliability Standard

    Science.gov (United States)

    2011-09-20

    ....C. Cir. 2009). \\4\\ Mandatory Reliability Standards for the Bulk-Power System, Order No. 693, FERC... for maintaining real and reactive power balance. \\14\\ Electric Reliability Organization Interpretation...; Order No. 753] Electric Reliability Organization Interpretation of Transmission Operations Reliability...

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  6. Colombeau's generalized functions and non-standard analysis

    International Nuclear Information System (INIS)

    Todorov, T.D.

    1987-10-01

    Using some methods of the Non-Standard Analysis we modify one of Colombeau's classes of generalized functions. As a result we define a class ε-circumflex of the so-called meta-functions which possesses all good properties of Colombeau's generalized functions, i.e. (i) ε-circumflex is an associative and commutative algebra over the system of the so-called complex meta-numbers C-circumflex; (ii) Every meta-function has partial derivatives of any order (which are meta-functions again); (iii) Every meta-function is integrable on any compact set of R n and the integral is a number from C-circumflex; (iv) ε-circumflex contains all tempered distributions S', i.e. S' is contained in ε' isomorphically with respect to all linear operations (including the differentiation). Thus, within the class ε-circumflex the problem of multiplication of the tempered distributions is satisfactorily solved (every two distributions in S' have a well-defined product in ε-circumflex). The crucial point is that C-circumflex is a field in contrast to the system of Colombeau's generalized numbers C-bar which is a ring only (C-bar is the counterpart of C-circumflex in Colombeau's theory). In this way we simplify and improve slightly the properties of the integral and notion of ''values of the meta-functions'' as well as the properties of the whole class ε-circumflex itself if compared with the original Colombeau theory. And, what is maybe more important, we clarify the connection between the Non-Standard Analysis and Colombeau's theory of new generalized functions in the framework of which the problem of multiplication of distributions was recently solved. (author). 14 refs

  7. Evolution of general design requirements for french pressurized water reactors

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Rollinger, F.

    1988-10-01

    The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5

  8. The Concept of the Use of the Marine Reactor Plant in Small Electric Grids

    International Nuclear Information System (INIS)

    Khlopkin, N.; Makarov, V.; Pologikh, B.

    2002-01-01

    In report some aspects of the using marine nuclear reactor are considered for provision of need small non-interconnected power systems, as well as separate settlements and the mining enterprises disposed in regions with a undeveloped infrastructure. Recently for these purposes it is offered to use the nuclear small modular power plants. The required plant power for small electric grids lies within from 1 to several tens of MWe. Module can be collected and tested on machine-building plant, and then delivered in ready type to the working place on some transport, for instance, a barge. Through determined time it's possible to transport a module to the repair shop and also to the point of storage after the end of operation. Marine nuclear reactors on their powers, compactness, mass and size are ideal prototypes for creation of such modules. For instance, building at present floating power unit, intended for functioning in region of the Russian North, based on using reactor plants of nuclear icebreakers. Reliability and safety of the ship reactor are confirmed by their trouble-free operation during approximately 180 reactors-years. Unlike big stationary nuclear plant, working in base mode, power unit with marine reactor wholly capable to work in mode of the loading following. In contrast with reactor of nuclear icebreaker, advisable to increase the core lifetime and to reduce the enrichment of the uranium. This requires more uranium capacity fuel compositions and design of the core. In particular, possible transition from traditional for ship reactor of the channel core to cassette design. Other directions of evolution of the ship reactors, not touching the basic constructive decisions verified by practice, but promoting development of properties of self-security of plant are possible. Among such directions is reduction volumetric power density of a core. (author)

  9. Study of reactor Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    1979-01-01

    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  10. Applications of nuclear reactor power systems to electric propulsion missions.

    Science.gov (United States)

    Schaupp, R. W.; Sawyer, C. D.

    1971-01-01

    The performance of nuclear electric propulsion systems (NEP) has been evaluated for a wide variety of missions in an attempt to establish the commonality of NEP system requirements. Emphasis was given to those requirements and system characteristics that serve as guidelines for current technology development programs. Various interactions and tradeoffs between NEP system and mission parameters are described. The results show that the most significant factors in selecting NEP system size are launch mode (direct or spiral escape) and, to a weaker extent, launch vehicle capability. Other factors such as mission, payload, and thrust time constraints, have little influence, thus allowing one NEP system to be used for many missions. The results indicated that a 100 kWe NEP would be suitable for most direct escape missions and a 250 kWe NEP system would be suitable for more demanding missions that use the spiral escape mode.

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  12. Expert systems for analysis and management assistance of nuclear reactor electrical power supplies

    International Nuclear Information System (INIS)

    Evrard, J.M.; Souchet, Y.

    1988-01-01

    Electrical power supplies of nuclear plants are very complex systems. They are studied in many ways: failure consequences, probabilistic risk assessement, failure diagnosis, corrective actions in case of incident ... Knowledge base technology (expert systems) is very suited to solve these problems. A common structural representation can generate specific functional representations; thus we get coherency and easy evolution. This paper shows a two facet methodology: the plant description (both static and dynamic) and reasoning about it. Current applications are developed for pressurized water reactors and gas cooled reactors, using the SPIRAL expert system shell buil at CEA [fr

  13. Device for providing a leak-tight penetration for electric cables through a reactor vault roof

    International Nuclear Information System (INIS)

    Eyral, M.; Mahe, A.

    1979-01-01

    The device for providing a cable penetration through the vault roof of a liquid sodium cooled fast reactor comprises a vertical tube closed at the top end by a flange-plate. Electric cables connected to measuring and detecting instruments are passed through the flange-plate which is joined to the reactor vault roof in leak-tight manner and enclosed within a removable hood. At least one horizontal plate is mounted within the vertical tube and provided with orifices for the leak-tight passage of the cables. Cable storage reels are placed within the tube and can be locked in position or released by controlled mechanical means

  14. Advanced marine reactor MRX and application to nuclear barge supplying electricity and heat

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Fukuhara, Yoshifumi; Ochiai, Masa-aki

    2000-01-01

    The basic design concept of an advanced marine reactor MRX has been established with adoption of several new technologies. The MRX is an integral-type PWR with 100 MWt aimed basically for use of ship propulsion. Adoption of a water-filled containment together with the integral type reactor makes the reactor light-weight and compact greatly. A engineered safety system is a simplified passive system, function of which is confirmed by the safety analysis. The MRX can be applied to an energy supply system of electricity and heat co-generation by installing it on a barge. Concept of a nuclear barge with the MRX of 334 MWt output is presented for use of supplying electricity, fresh water and hot water. Combined system of electric generation and desalination with the RO process can deliver variable output of electricity and fresh water according a demand. Latent heat of the exhausted steam from the turbine can be used effectively to raise the temperature of cold water as heat supply. (author)

  15. 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)

  16. An inherently safe power reactor module

    International Nuclear Information System (INIS)

    Salerno, L.N.

    1985-01-01

    General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

  17. Analysis of the General Electric Company swell tests with RELAP4/MOD7

    International Nuclear Information System (INIS)

    Fischer, S.R.; Hendrix, C.E.

    1979-01-01

    The RELAP4/MOD7 nuclear reactor transient analysis code, presently being developed by EG and G Idaho, Inc., will incorporate several significant improvements over earlier versions of RELAP4. As part of the development of RELAP4/MOD7, a thorough assessment of the capability of the code to simulate water reactor LOCA phenomena is being made. This assessment is accomplished in part by comparing results from code calculations with test data from experimental facilities. Simulations of the General Electric Company (GE) level swell tests were performed as part of the code checkout. In these tests, a pressurized vessel partially filled with nearly saturated water was blown down through a simulated break located near the top of the vessel. Comparison of RELAP4 calculations with data from these experiments indicates that the code has the capability to model the unequal phase velocity flow and resulting density gradients that might occur in a BWR steam line break transient. Comparisons of RELAP4 calculations with data from two level swell experiments are presented

  18. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.

  19. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    International Nuclear Information System (INIS)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed

  20. 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

  1. 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

  2. Standard technical specifications for combustion engineering pressurized water reactors

    International Nuclear Information System (INIS)

    1979-08-01

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

  3. 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

  4. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  5. Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

    International Nuclear Information System (INIS)

    Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

    1982-01-01

    Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

  6. Radioactive waste shipments to Hanford Retrievable Storage from the General Electric Vallecitos Nuclear Center, Pleasanton, California

    International Nuclear Information System (INIS)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I.; Duncan, D.R.

    1993-10-01

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Approximately 3.8% of the TRU waste to be retrieved for shipment to WIPP was generated at the General Electric (GE) Vallecitos Nuclear Center (VNC) in Pleasanton, California and shipped to the Hanford Site for storage. The purpose of this report is to characterize these radioactive solid wastes using process knowledge, existing records, and oral history interviews. The waste was generated almost exclusively from the activities, of the Plutonium Fuels Development Laboratory and the Plutonium Analytical Laboratory. Section 2.0 provides further details of the VNC physical plant, facility operations, facility history, and current status. The solid radioactive wastes were associated with two US Atomic Energy Commission/US Department of Energy reactor programs -- the Fast Ceramic Reactor (FCR) program, and the Fast Flux Test Reactor (FFTR) program. These programs involved the fabrication and testing of fuel assemblies that utilized plutonium in an oxide form. The types and estimated quantities of waste resulting from these programs are discussed in detail in Section 3.0. A detailed discussion of the packaging and handling procedures used for the VNC radioactive wastes shipped to the Hanford Site is provided in Section 4.0. Section 5.0 provides an in-depth look at this waste including the following: weight and volume of the waste, container types and numbers, physical description of the waste, radiological components, hazardous constituents, and current storage/disposal locations

  7. Radioactive waste shipments to Hanford Retrievable Storage from the General Electric Vallecitos Nuclear Center, Pleasanton, California

    Energy Technology Data Exchange (ETDEWEB)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I. [Los Alamos Technical Associates, Inc., NM (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-10-01

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Approximately 3.8% of the TRU waste to be retrieved for shipment to WIPP was generated at the General Electric (GE) Vallecitos Nuclear Center (VNC) in Pleasanton, California and shipped to the Hanford Site for storage. The purpose of this report is to characterize these radioactive solid wastes using process knowledge, existing records, and oral history interviews. The waste was generated almost exclusively from the activities, of the Plutonium Fuels Development Laboratory and the Plutonium Analytical Laboratory. Section 2.0 provides further details of the VNC physical plant, facility operations, facility history, and current status. The solid radioactive wastes were associated with two US Atomic Energy Commission/US Department of Energy reactor programs -- the Fast Ceramic Reactor (FCR) program, and the Fast Flux Test Reactor (FFTR) program. These programs involved the fabrication and testing of fuel assemblies that utilized plutonium in an oxide form. The types and estimated quantities of waste resulting from these programs are discussed in detail in Section 3.0. A detailed discussion of the packaging and handling procedures used for the VNC radioactive wastes shipped to the Hanford Site is provided in Section 4.0. Section 5.0 provides an in-depth look at this waste including the following: weight and volume of the waste, container types and numbers, physical description of the waste, radiological components, hazardous constituents, and current storage/disposal locations.

  8. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    Blom, K.H.; Krull, W.

    1997-01-01

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  10. Design study of electrical power supply system for tokamak fusion power reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)

  11. The roles of ozone and zeolite on reactive dye degradation in electrical discharge reactors.

    Science.gov (United States)

    Peternel, L; Kusic, H; Koprivanac, N; Locke, B R

    2006-05-01

    In this study high voltage pulsed corona electrical discharge advanced oxidation processes (AOPs) were applied to bleach and degrade C.I. Reactive Green 8 and C.I. Reactive Red 45 organic dyes in water solutions. Two types of hybrid gas/liquid high voltage electrical discharge (corona) reactors, known as hybrid series and hybrid parallel were studied. The difference between these reactors relates to electrode configuration, which affects the amounts of ozone, hydrogen peroxide and hydroxyl radicals produced. Experiments were conducted using dye concentrations of 20 mgl(-1) and 75 mgl(-1), with and without NH4ZSM5 zeolite addition in order to determine possible effects of added solid particles to total process efficiency. The role of ozone in combination with zeolites was assessed through comparative direct ozonation experiments with ozone supplied by an ozone generator. UV/VIS spectrophotometric measurements and measurements of total organic carbon (TOC) were used for the determination of decolorization and mineralization rates.

  12. Small nuclear power reactor emergency electric power supply system reliability comparative analysis

    International Nuclear Information System (INIS)

    Bonfietti, Gerson

    2003-01-01

    This work presents an analysis of the reliability of the emergency power supply system, of a small size nuclear power reactor. Three different configurations are investigated and their reliability analyzed. The fault tree method is used as the main tool of analysis. The work includes a bibliographic review of emergency diesel generator reliability and a discussion of the design requirements applicable to emergency electrical systems. The influence of common cause failure influences is considered using the beta factor model. The operator action is considered using human failure probabilities. A parametric analysis shows the strong dependence between the reactor safety and the loss of offsite electric power supply. It is also shown that common cause failures can be a major contributor to the system reliability. (author)

  13. 76 FR 3517 - Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial...

    Science.gov (United States)

    2011-01-20

    ... Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial-Institutional, and... following: Category NAICS \\1\\ Examples of regulated entities Industry 221112 Fossil fuel-fired electric utility steam generating units. Federal Government 22112 Fossil fuel-fired electric utility steam...

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

    Science.gov (United States)

    2011-03-16

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

  15. DESIGN SAFETY PROBLEMS OF NUCLEAR REACTORS IN SPACE FOR ELECTRICAL POWER

    Energy Technology Data Exchange (ETDEWEB)

    Pickler, D A

    1963-06-15

    A general treatment is presented of some of the problems in the design safety of reactors which are to be operated in space. The basic requirements of these reachigh temperatures. The usual concept of a space reactor is described briefly, and the hazards of an assumed unmanned vehicle with an enriched-U-fueled reactor are examined during its launching, orbit, and reentry. Graphs are given for the dose vs distance downwind for an excursion of 100 Mw-sec, for the activity vs time after shutdown of a reactor which has been operated for 5 yr at 100 kw(t), and for the altitude vs orbital lifetime. Apparent conflicts between the basic requirements are discussed. (D.L.C.)

  16. 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

  17. 78 FR 41907 - Effluent Limitations Guidelines and Standards for the Steam Electric Power Generating Point...

    Science.gov (United States)

    2013-07-12

    ... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 423 [EPA-HQ-OW-2009-0819. FRL-9832-7; EPA-HQ-RCRA-2013-0209] RIN 2040-AF14 Effluent Limitations Guidelines and Standards for the Steam Electric Power... proposed rule entitled, ``Effluent Limitations Guidelines and Standards for the Steam Electric Power...

  18. A generalization information management system applied to electrical distribution

    Energy Technology Data Exchange (ETDEWEB)

    Geisler, K.I.; Neumann, S.A.; Nielsen, T.D.; Bower, P.K. (Empros Systems International (US)); Hughes, B.A.

    1990-07-01

    This article presents a system solution approach that meets the requirements being imposed by industry trends and the electric utility customer. Specifically, the solution addresses electric distribution management systems. Electrical distribution management is a particularly well suited area of application because it involves a high diversity of tasks, which are currently supported by a proliferation of automated islands. Islands of automation which currently exist include (among others) distribution operations, load management, automated mapping, facility management, work order processing, and planning.

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

    International Nuclear Information System (INIS)

    Misak, J.; Rohar, S.

    1995-01-01

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

  20. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  1. 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...

  2. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    International Nuclear Information System (INIS)

    Touati, Said; Chennai, Salim; Souli, Aissa

    2015-01-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  3. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    Energy Technology Data Exchange (ETDEWEB)

    Touati, Said; Chennai, Salim; Souli, Aissa [Nuclear Research Centre of Birine, Ain Oussera, Djelfa Province (Algeria)

    2015-07-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  4. Technical regulations on the general design and safety criteria for design and construction of nuclear reactors of May 1975

    International Nuclear Information System (INIS)

    1975-05-01

    These Technical Regulations published on 5th September 1975 were made in implementation of Section 33 of Decree No 7/9141 on the procedure for the licensing of nuclear installations. They serve as a guide to licensing authorities, project designers and operators in the nuclear field and therefore provide general criteria for safety standards, engineering codes, siting considerations, design bases for overall environmental radiation protection, and also deal with reactor core design, instrumentation, control, alarm systems, including an emergency core cooling system. Finally, the safe design of fuel elements must be ensured and fuel storage and handling techniques complied with. (NEA) [fr

  5. Study of design parameters for minimizing the cost of electricity of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Tokimatsu, K.; Yamaji, K.; Katsurai, M.; Okano, K.; Yoshida, T.

    1998-01-01

    The impact of the design parameters on the cost of electricity (COE) is studied through a parameter survey in order to minimize the COE. Three kinds of operating modes are considered; first stability (FS), second stability (SS) and reversed shear (RS). The COE is calculated by a coupled physics-engineering-cost computer system code. Deuterium-tritium type, 1000 MW(e) at electric bus bar, steady state tokamak reactors with aspect ratios A from 3 to 4.5 are assumed. Several criteria are used for the parameter survey; for example, (a) the thermal to electrical conversion efficiency is assumed to be 34.5% using water as a coolant; (b) the average neutron wall load must not exceed 5 MW/m 2 for plasma major radius R p >5 m; (c) a 2 MeV neutral beam injector (NBI) is applied. It is found that the RS operating mode most minimizes the COE among the three operating modes by reducing the cost of the current drive and the coils and structures. The cost-minimized RS reactor can attain high f bs , high β N and low q 95 at the same time, which results in a short R p of 5.1 m, a low B max of the maximum magnetic toroidal field (TF) of the TF coils of 13 T and a low A of 3.0. It can be concluded that this cost-minimized RS reactor is the most cost-minimized within the frameworks of this study. This cost-minimized RS reactor has two advantages: one is that a B max =13 T TF coil can be made by use of ITER coil technology and the other is that the same cooling technology as that of ITER (water cooling) can be used. (author)

  6. Performance of the Lead-Alloy-Cooled Reactor Concept Balanced for Actinide Burning and Electricity Production

    International Nuclear Information System (INIS)

    Hejzlar, Pavel; Davis, Cliff B.

    2004-01-01

    A lead-bismuth-cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners

  7. Online calibration method for condition monitoring of nuclear reactor instrumentations based on electrical signature analysis

    International Nuclear Information System (INIS)

    Syaiful Bakhri

    2013-01-01

    Electrical signature analysis currently becomes an alternative in condition monitoring in nuclear power plants not only for stationary components such as sensors, measurement and instrumentation channels, and other components but also for dynamic components such as electric motors, pumps, generator or actuators. In order to guarantee the accuracy, the calibration of monitoring system is a necessary which practically is performed offline, under limited schedules and certain tight procedures. This research aims to introduce online calibration technique for electrical signature condition monitoring in order that the accuracy can be maintained continuously which in turn increases the reactor safety as a whole. The research was performed step by stepin detail from the conventional technique, online calibration using baseline information and online calibration using differential gain adjustment. Online calibration based on differential gain adjustment provides better results than other techniques even tough under extreme gain insertion as well as external disturbances such as supply voltages. (author)

  8. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  9. A small-scale modular reactor for electric source for remote places

    International Nuclear Information System (INIS)

    Anon.

    2002-01-01

    Use of a small-scale modular reactor (SMR) as an electric source for remote places is one of scenarios for actual use of SMR parallel to alternative source of present nuclear power stations and co-generation source at urban suburbs, there is not only an actual experience to construct and operate for power source for military use in U.S.A. on 1950s to 1960s, but also four nuclear reactors (LWGR, 12 MW) in Vilyvino Nuclear Power Station in far northern district in Russia are under operation. Recently, Department of Energy in U.S.A. prepared the 'Report to Congress on Small Modular Nuclear Reactors' evaluating on feasibility of SMR as a power source for remote places according to requirement of the Congress. This report evaluated a feasibility study on nine SMRs in the world with 10 to 50 MW of output as electric source for remote places on economical efficiency and so on, together with analysis of their design concepts, to conclude that 'they could perform beginning of operations on 2000s because of no large technical problems and keeping a level capable of competing with power generation cost at remote place on its present economical efficiency'. Here was introduced on outlines of this report. (G.K.)

  10. The role of standardization for innovation in the electric car inudstry

    DEFF Research Database (Denmark)

    Andersen, Frank Steven Dahl

    This paper investigates the role of standardization for innovation among selected actors in the electric car industry. The case indicates that both formal standards and radical innovations have been developed, and that standards in themself do not exclude the existence or development of innovations......; in fact standards may contribute positively to innovation. The essential standards developed so far have mainly brought compatibility among components and innovations in the industry, which made it possible to electrical car manufacturers to align innovation activities and adapt new technology effectively...

  11. 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

  12. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-02-01

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out for several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 365 MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule, and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating, or desalination

  13. Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

    International Nuclear Information System (INIS)

    Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

    1978-01-01

    The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

  14. Oxygen transport membrane reactor based method and system for generating electric power

    Science.gov (United States)

    Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan

    2017-02-07

    A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

  15. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-01-01

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313 MW(t) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(t), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(t) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination

  16. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-01-01

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination. (author)

  17. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

    1999-01-01

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  18. Basic Guidelines to Introduce Electric Circuit Simulation Software in a General Physics Course

    Science.gov (United States)

    Moya, A. A.

    2018-01-01

    The introduction of electric circuit simulation software for undergraduate students in a general physics course is proposed in order to contribute to the constructive learning of electric circuit theory. This work focuses on the lab exercises based on dc, transient and ac analysis in electric circuits found in introductory physics courses, and…

  19. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF... Denmark Finland France Germany Greece Indonesia Ireland Italy Japan Latvia Lithuania Luxembourg...

  20. SACRD: a data base for fast reactor safety computer codes, general description

    International Nuclear Information System (INIS)

    Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Simpson, D.B.; Flanagan, G.F.

    1979-01-01

    SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics, and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases. A general description of the SACRD system is presented in the report

  1. Utilization of nuclear energy for generating electric power in the FRG, with special regard to LWR-type reactors

    International Nuclear Information System (INIS)

    Vollradt, J.

    1977-01-01

    Comments on interdependencies in energy industry and energy generation as seen by energy supply utilities, stating that the generation of electric power in Germany can only be based on coal and nuclear energy in the long run, are followed by the most important, fundamental, nuclear-physical, technological and in part political interdependencies prevailing in the starting situation of 1955/58 when the construction of nuclear power plant reactors began. Then the development ranging to the 28000 MW nuclear power output to be expected in 1985 is outlined, totalling in 115000 MW electric power in the FRG. Finally, using the respectively latest order, the technical set up of each of the reactor types with 1300 MWe unit power offered by German manufacturers are described: BBC/BBR PWR-type reactor Neupotz, KWU-PWR-type reactor Hamm and KWU PWR-type reactor double unit B+C Gundremmingen. (orig.) [de

  2. 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.

  3. 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.)

  4. GT-MHR as economical highly efficient inherently safe modular gas cooled reactor for electric power generation

    International Nuclear Information System (INIS)

    Baydakov, A.G.; Kodochigov, N.G.; Kuzavkov, N.G.; Vorontsov, V.E.

    2001-01-01

    The Gas Turbine Modular Helium Reactor (GT-MHR) project is a joint effort of organisations in the Russian Federation co-ordinated by the Ministry on Atomic Energy (MINATOM), General Atomics in the United States, Framatome in France, and Fuji Electric in Japan. This paper describes the structure, objectives and history of the GT-MHR project, as well as the plant design, including plant layout, components and fuel. It summarises the results of evaluations of the design safety characteristics and economics, for both a plutonium fueled single module and a uranium fueled four module plant. The plans and anticipated structure for an international project to construct the plant are also summarised. (author)

  5. Propagation of negative electrical discharges through 2-dimensional packed bed reactors

    International Nuclear Information System (INIS)

    Kruszelnicki, Juliusz; Engeling, Kenneth W; Foster, John E; Xiong, Zhongmin; Kushner, Mark J

    2017-01-01

    Plasma-based pollutant remediation and value-added gas production have recently gained increased attention as possible alternatives to the currently-deployed chemical reactor systems. Electrical discharges in packed bed reactors (PBRs) are of interest, due to their ability to synergistically combine catalytic and plasma chemical processes. In principle, these systems could be tuned to produce specific products, based on their application by combinations of power formats, materials, geometries and working gases. Negative voltage, atmospheric-pressure plasma discharges sustained in humid air in a PBR-like geometry were experimentally characterized using ICCD imaging and simulated in 2-dimensions (2D) to provide insights into possible routes to this tunability. Surface ionization waves (SIWs) and positive restrikes through the lattice of dielectric rods were shown to be the principal means of producing reactive species. The number and intensity of SIWs and restrikes are sensitive functions of the alignment of the lattice of dielectric beads (or rods in 2D) with respect to the applied electric field. Decreased spacing between the dielectric elements leads to an increased electric field enhancement in the gas, and therefore locally higher plasma densities, but does not necessarily impact the types of discharges that occur through the lattice. (paper)

  6. Study on the profitableness of electricity generation with high temperature reactors

    International Nuclear Information System (INIS)

    Kolb, G.

    1978-08-01

    The programme group 'Systemforschung und Technologische Entwicklung' (STE) of the Nuclear Research Centre Juelich in cooperation with the internal institutions 'Projekttraegerschaft Entwicklung von Hochtemperaturreaktoren' (PTH) and 'Institut fuer Reaktorentwicklung' (IRE) on the one hand, and the external partner 'Hochtemperatur-Reactorbau GmbH' (HRB) and 'Gesellschaft fuer Hochtemperaturreactor Technik' (GHT) on the other hand have set up a study on fuel cycle costs, electricity production cost and the economical use as well as uranium resource protection by introduction of high temperature reactors (HTR) with pebble bed core to generate electricity. The pressurized-water reactor (PWR) today on the market serves as comparison. The working results obtained sofar are compiled in the present report. It was particularly noted that - the HTR can economically fully compete with the PWR for electricity generation - the necessary supply of natural uranium for the HTR in open circuit is about one third lower and in the closed circuit, almost two thirds lower than in the corresponding PWR. A further reduction is possible on a long-term basis by highly converting HTW systems. (orig.) [de

  7. Propagation of negative electrical discharges through 2-dimensional packed bed reactors

    Science.gov (United States)

    Kruszelnicki, Juliusz; Engeling, Kenneth W.; Foster, John E.; Xiong, Zhongmin; Kushner, Mark J.

    2017-01-01

    Plasma-based pollutant remediation and value-added gas production have recently gained increased attention as possible alternatives to the currently-deployed chemical reactor systems. Electrical discharges in packed bed reactors (PBRs) are of interest, due to their ability to synergistically combine catalytic and plasma chemical processes. In principle, these systems could be tuned to produce specific products, based on their application by combinations of power formats, materials, geometries and working gases. Negative voltage, atmospheric-pressure plasma discharges sustained in humid air in a PBR-like geometry were experimentally characterized using ICCD imaging and simulated in 2-dimensions (2D) to provide insights into possible routes to this tunability. Surface ionization waves (SIWs) and positive restrikes through the lattice of dielectric rods were shown to be the principal means of producing reactive species. The number and intensity of SIWs and restrikes are sensitive functions of the alignment of the lattice of dielectric beads (or rods in 2D) with respect to the applied electric field. Decreased spacing between the dielectric elements leads to an increased electric field enhancement in the gas, and therefore locally higher plasma densities, but does not necessarily impact the types of discharges that occur through the lattice.

  8. Fusion neutronics plan in the development of fusion reactor. With the aim of realizing electric power

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hiroo; Morimoto, Yuichi; Ochiai, Kentarou; Sugimoto, Masayoshi; Nishitani, Takeo; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    On June 1992, Atomic Energy Commission in Japan has settled Third Phase Program of Fusion Research and Development to achieve self-ignition condition, to realize long pulse burning plasma and to establish basis of fusion engineering for demonstration reactor. This report describes research plan of Fusion Neutron Laboratory in JAERI toward a development of fusion reactor with an aim of realizing electric power. The fusion neutron laboratory has a fusion neutronics facility (FNS), intense fusion neutron source. The plan includes research items in the FNS; characteristics of shielding and breeding materials, nuclear characteristics of materials, fundamental irradiation process of insulator, diagnostics materials and structural materials, and development of in-vessel diagnostic technology. Upgrade of the FNS is also described. Also, the International Fusion Material Irradiation Facility (IFMIF) for intense neutron source to develop fusion materials is described. (author)

  9. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  10. Recovery of reactor electrical assemblies using differential de-encapsulation to remove dielectric insulation systems

    International Nuclear Information System (INIS)

    Hubrig, J.G.; Hammerstone, E.B.

    1986-01-01

    State-of-the-art de-encapsulation technologies associated with the conventional dielectric insulation systems employed in the construction of electrical coils and power distribution systems do not allow for accurate fatigue/failure analysis or reliable recovery of costly assembly components. Differential de-encapsulation allows for the selective removal of contemporary thermoset resin based insulation systems to allow non-destructive penetration of insulation wall thicknesses to both examine critical areas and recover high performance metallic and non-metallic inserts for remanufacture; significantly reducing replacement costs and reactor downtime. The authors' analysis describes how the availability of engineering data from the selective and non-destructive removal of insulation materials will aid in the evaluation of original manufacture, materials and procedures; enabling redesign to enhance subsequent on line performance. They also discuss why the ability to recover coil and core assemblies for remanufacture will have a major economic impact on reactor management costs

  11. Analysis and evaluation of ZPPR critical experiments for a 100 kilowatt-electric space reactor

    International Nuclear Information System (INIS)

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W.; Doncals, R.A.; Andre, S.V.; Porter, C.A.; Cowan, C.L.; Stewart, S.L.; Protsik, R.

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously needed for fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further reductions in the mass of the SP-100. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design

  12. The uses and implications of standards in general practice consultations

    DEFF Research Database (Denmark)

    Lippert, Maria Laura; Reventlow, Susanne; Kousgaard, Marius Brostrøm

    2017-01-01

    Quality standards play an increasingly important role in primary care through their inscription in various technologies for improving professional practice. While ‘hard’ biomedical standards have been the most common and debated, current quality development initiatives increasingly seek to include...... as manifestations of an inherent conflict between principles of patient-centredness and formal biomedical quality standards. However, this study suggests that standards on the ‘softer’ aspects of care may just as well interfere with a clinical approach relying on situated and attentive interactions with patients....

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to January 1981

  15. Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

    International Nuclear Information System (INIS)

    Liu Kukui; Li Manchang; Tang Chuanbao

    1997-01-01

    In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

  16. Development of dust removal system using static electricity for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori; Ueda, Yasutoshi; Oda, Yasushi; Takahashi, Kenji [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao; Ueda, Shuzo; Kurihara, Ryoichi

    1997-11-01

    Tests to collect and transport metallic and non-metallic dust particles have been conducted using static electricity in a vacuum environment to investigate the applicability of a static electricity dust removal system for fusion experimental reactors. The dust particles are charged by electrostatic induction, floated and collected due to the Coulomb force generated by the AC electric field. They are then transported due to the gradient force induced by the electric curtain of the non-uniform travelling-wave electric field. Using a fully insulated electrode with a single-phase AC voltage up to 15 kV, aluminum and carbon dust were successfully collected. The highest collection rates for the aluminum and carbon dust were around 30 and 2 g/min, respectively. The linear-type electrodes, using as high as 22 kV of the three-phase AC voltage, transported aluminum dust up to an angle of 60deg. Applying a guide electrode to the linear-type electrode, the transportation rate was approximately doubled and almost constant at every angle, including a 90deg angle. The system transported aluminum dust up to the rate of 13 g/min. The influence of the 0.15 T magnetic field on the dust collection and transportation efficiencies was found to be negligible. (author)

  17. Development of dust removal system using static electricity for fusion experimental reactors

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Ueda, Yasutoshi; Oda, Yasushi; Takahashi, Kenji; Seki, Yasushi; Aoki, Isao; Ueda, Shuzo; Kurihara, Ryoichi.

    1997-01-01

    Tests to collect and transport metallic and non-metallic dust particles have been conducted using static electricity in a vacuum environment to investigate the applicability of a static electricity dust removal system for fusion experimental reactors. The dust particles are charged by electrostatic induction, floated and collected due to the Coulomb force generated by the AC electric field. They are then transported due to the gradient force induced by the electric curtain of the non-uniform travelling-wave electric field. Using a fully insulated electrode with a single-phase AC voltage up to 15 kV, aluminum and carbon dust were successfully collected. The highest collection rates for the aluminum and carbon dust were around 30 and 2 g/min, respectively. The linear-type electrodes, using as high as 22 kV of the three-phase AC voltage, transported aluminum dust up to an angle of 60deg. Applying a guide electrode to the linear-type electrode, the transportation rate was approximately doubled and almost constant at every angle, including a 90deg angle. The system transported aluminum dust up to the rate of 13 g/min. The influence of the 0.15 T magnetic field on the dust collection and transportation efficiencies was found to be negligible. (author)

  18. Use of small reactors as an alternative to supply electricity to Baja California Sur; Uso de reactores pequenos como alternativa de suministro de electricidad para Baja California Sur

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, G.; Portes, E.; Ramirez, J. R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ortega, G., E-mail: gustavo.alonso@inin.gob.mx [Comision Federal de Electricidad, Rio Rodano No. 14, 06500 Ciudad de Mexico (Mexico)

    2016-09-15

    The state of Baja California Sur (Mexico) does not form part of the national interconnected electrical system of the country, reason why is local its electrical power supply; one of the alternatives to cover future demands is the use of gas-based combined cycles, which presents the additional problem of including a high price for gas transportation in its costs. In order to reduce total costs, including investment, fuels and operation and maintenance in the operation of the Baja California Sur state electricity system in the coming years, mainly due to the estimated natural gas cost order of $11.50 dollars per million BTU, a proposal is presented to reduce the costs of the electrical system by replacing the necessary combined cycles with the new Small Modular Reactor type nuclear reactors, this alternative is economically competitive. (Author)

  19. 40 CFR 94.7 - General standards and requirements.

    Science.gov (United States)

    2010-07-01

    ... to the emission standards of this part are equipped with a connection in the engine exhaust system that is located downstream of the engine and before any point at which the exhaust contacts water (or... be internally threaded with standard pipe threads of a size not larger than one-half inch, and shall...

  20. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs

  1. General plan for the partial dismantling of the IRT-Sofia research reactor

    Directory of Open Access Journals (Sweden)

    Apostolov Tihomir G.

    2006-01-01

    Full Text Available After the decision of the Bulgarian Government to reconstruct it, the strategy concerning the IRT-Sofia Research Reactor is to partially dismantle the old systems and equipment. The removal of the reactor core and replacement of old equipment will not pose any significant problems. For a more efficient use of existing resources, there is a need for an engineering project which has been already prepared under the title "General Plan for the Partial Dismantling of Equipment at the IRT-Sofia as a Part of the Reconstruction into a Low Power RR".

  2. Alteration in reactor installation (addition of Unit 2) in Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc. (inquiry)

    International Nuclear Information System (INIS)

    1983-01-01

    An inquiry was made by the Ministry of International Trade and Industry to Nuclear Safety Commission on the addition of Unit 2 in Shimane Nuclear Power Station of The Chugoku Electric Power Co., Inc., concerning the technical capability of Chugoku Electric Power Co., Inc., and the plant safety. The NSC requested the Committee on Examination of Reactor Safety to make a deliberation on this subject. Both the technical capability and the safety of Unit 1 were already confirmed by MITI. Unit 2 to be newly added in the Shimane Nuclear Power Station is a BWR power plant with electric output of 820 MW. The examination made by MITI is described: the technical capability of Chugoku Electric Power Co., Inc., the safety of Unit 2 about its siting, reactor proper, reactor cooling system, radioactive waste management, etc. (J.P.N.)

  3. 78 FR 44899 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-07-25

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Notice of proposed... Electric Company (GE) GE90-110B1 and -115B turbofan engines. This proposed AD was prompted by multiple...) 2165M22P01, installed on GE90-110B1 and -115B turbofan engines. One of the leaks led to an under cowl engine...

  4. 78 FR 19628 - Airworthiness Directives; General Electric Company Turbofan Engines

    Science.gov (United States)

    2013-04-02

    ... Company Turbofan Engines AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Notice of proposed... Electric Company (GE) GE90-76B, -85B, -90B, -94B, - 110B1, and -115B turbofan engines. This proposed AD was... of stage 1 HPT stator shroud distress resulting in engine removals on airplanes with GE90 turbofan...

  5. Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M., E-mail: guimarae@ieav.cta.br, E-mail: lamartine.guimaraes@pq.cnpq.br, E-mail: jamil@ieav.cta.br, E-mail: jalnsgf@outlook.com, E-mail: borges.em@hotmail.com, E-mail: ecorborges@hotmail.com, E-mail: ivayolini@gmail.com, E-mail: guilherme_placco@ig.com.br [Instituto de Estudos Avancados (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil); Barrios Junior, Ary Garcia, E-mail: arygarcia89@yahoo.com [Faculdade de Tecnologia Sao Francisco (FATESF), Jacarei, SP (Brazil)

    2013-07-01

    This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

  6. Development of electrically insulating self-healing coatings in vanadium alloys for lithium fusion reactor

    International Nuclear Information System (INIS)

    1999-01-01

    Problems on electrically insulating self-healing coatings (SHC) on vanadium alloys for lithium fusion reactor systems are considered. In particular, the SHC stability and radiation resistance in lithium and effect of magnetic field on the efficiency of the TNR lithium systems are studied. New technological methods for application of self-healing coatings and study on their properties are developed. The vanadium-lithium materials testing in pile loops for solution of the above problems under conditions of the lithium TNR is described [ru

  7. Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M.; Barrios Junior, Ary Garcia

    2013-01-01

    This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

  8. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre

    International Nuclear Information System (INIS)

    2007-05-01

    This note of information resumes the progress of the electric failure on the reactor n.3 of the nuclear power plant of Dampierre, the organization during the incident, it establishes then a comparison with the incident arisen to Forsmark in 2006 and reminds that it lead in an inspection on behalf of the Asn which noticed that all the procedures had been respected by the operators and did not noticed any abnormality in the maintenance. This event was classified at the level 1 of the international nuclear event scale (INES). (N.C.)

  9. 78 FR 2797 - Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and Electric Vehicles

    Science.gov (United States)

    2013-01-14

    ... Sound Requirements for Hybrid and Electric Vehicles; Draft Environmental Assessment for Rulemaking To Establish Minimum Sound Requirements for Hybrid and Electric Vehicles; Proposed Rules #0;#0;Federal Register...-0148] RIN 2127-AK93 Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and...

  10. Can British Columbia Achieve Electricity Self-Sufficiency and Meet its Renewable Portfolio Standard?

    NARCIS (Netherlands)

    Sopinka, A.; Kooten, van G.C.; Wong, L.

    2012-01-01

    British Columbia’s energy policy is at a crossroads; the province has set a goal of electricity self-sufficiency, a 93% renewable portfolio standard and provincial natural gas strategy that could increase electricity consumption by 2,500-3,800 MW. To ascertain the reality of BC’s supply position, we

  11. The Nuclear option for U.S. electrical generating capacity additions utilizing boiling water reactor technology

    International Nuclear Information System (INIS)

    Garrity, T.F.; Wilkins, D.R.

    1993-01-01

    The technology status of the Advanced Boiling Water (ABWR) and Simplified Boiling Water (SBWR) reactors are presented along with an analysis of the economic potential of advanced nuclear power generation systems based on BWR technology to meet the projected domestic electrical generating capacity need through 2005. The forecasted capacity needs are determined for each domestic North American Electric Reliability Council (NERC) region. Extensive data sets detailing each NERC region's specific generation and load characteristics, and capital and fuel cost parameters are utilized in the economic analysis of the optimal generation additions to meet this need by use of an expansion planning model. In addition to a reference case, several sensitivity cases are performed with regard to capital costs and fuel price escalation

  12. 29 CFR 1607.5 - General standards for validity studies.

    Science.gov (United States)

    2010-07-01

    ...) (hereinafter “A.P.A. Standards”) and standard textbooks and journals in the field of personnel selection. D... market and the job should be considered in the determination of when a validity study is outdated. ...

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

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  15. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-01-01

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal

  16. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  17. A novel Generalized State-Space Averaging (GSSA) model for advanced aircraft electric power systems

    International Nuclear Information System (INIS)

    Ebrahimi, Hadi; El-Kishky, Hassan

    2015-01-01

    Highlights: • A study model is developed for aircraft electric power systems. • A novel GSSA model is developed for the interconnected power grid. • The system’s dynamics are characterized under various conditions. • The averaged results are compared and verified with the actual model. • The obtained measured values are validated with available aircraft standards. - Abstract: The growing complexity of Advanced Aircraft Electric Power Systems (AAEPS) has made conventional state-space averaging models inadequate for systems analysis and characterization. This paper presents a novel Generalized State-Space Averaging (GSSA) model for the system analysis, control and characterization of AAEPS. The primary objective of this paper is to introduce a mathematically elegant and computationally simple model to copy the AAEPS behavior at the critical nodes of the electric grid. Also, to reduce some or all of the drawbacks (complexity, cost, simulation time…, etc) associated with sensor-based monitoring and computer aided design software simulations popularly used for AAEPS characterization. It is shown in this paper that the GSSA approach overcomes the limitations of the conventional state-space averaging method, which fails to predict the behavior of AC signals in a circuit analysis. Unlike conventional averaging method, the GSSA model presented in this paper includes both DC and AC components. This would capture the key dynamic and steady-state characteristics of the aircraft electric systems. The developed model is then examined for the aircraft system’s visualization and accuracy of computation under different loading scenarios. Through several case studies, the applicability and effectiveness of the GSSA method is verified by comparing to the actual real-time simulation model obtained from Powersim 9 (PSIM9) software environment. The simulations results represent voltage, current and load power at the major nodes of the AAEPS. It has been demonstrated that

  18. Photovoltaic electricity pending the emergence of a new technological standard

    International Nuclear Information System (INIS)

    Tchinda, R.

    1999-01-01

    After a general overview of the policy and the institutional and legal framework existing for new and renewable energy in Cameroon, the authors focus on the potential of each of the sources, detailing the problems encountered with the operation of each of them in turn. They then consider the problems to be encountered in this sub-sector for new and renewable energy, and propose actions to be undertaken to provide solutions. (author)

  19. Conceptual design of a nucleo electric simulator with PBMR reactor based in Reduced order models

    International Nuclear Information System (INIS)

    Valle H, J.; Morales S, J.B.

    2005-01-01

    This project has as purpose to know to depth the operation of a PBMR nucleo electric type (Pebble Bed Modular Reactor), which has a reactor of moderate graphite spheres and fuel of uranium dioxide cooled with Helium and Brayton thermodynamic cycle. The simulator seeks to describe the dynamics of the one process of energy generation in the nuclear fuel, the process of transport toward the coolant one and the conversion to mechanical energy in the turbo-generators as well as in the heat exchangers indispensable for the process. The dynamics of reload of the fuel elements it is not modeled in detail but their effects are represented in the parameters of the pattern. They are modeled also the turbo-compressors of the primary circuit of the work fluid. The control of the power of the nuclear reactor is modeled by means of reactivity functions specified in the simulation platform. The proposed mathematical models will be settled in the platform of simulation of Simulink-Mat Lab. The proposed control panels for this simulator can be designed and to implement using the box of tools of Simulink that facilitates this process. The work presents the mathematical models more important used for their future implementation in Simulink. (Author)

  20. Potential minimum cost of electricity of superconducting coil tokamak power reactors

    International Nuclear Information System (INIS)

    Reid, R.L.; Peng, Y-K. M.

    1989-01-01

    The potential minimum cost of electricity (COE) for superconducting tokamak power reactors is estimated by increasing the physics (confinement, beta limit, bootstrap current fraction) and technology [neutral beam energy, toroidal field (TF) coil allowable stresses, divertor heat flux, superconducting coil critical field, critical temperature, and quench temperature rise] constraints far beyond those assumed for ITER until the point of diminishing returns is reached. A version of the TETRA systems code, calibrated with the ITER design and modified for power reactors, is used for this analysis, limiting this study to reactors with the same basic device configuration and costing algorithms as ITER. A minimum COE is reduced from >200 to about 80 mill/kWh when the allowable design constraints are raised to 2 times those of ITER. At 4 times the ITER allowables, a minimum COE of about 60 mill/kWh is obtained. The corresponding tokamak has a major radius of approximately 4 m, a plasma current close to 10 MA, an aspect ratio of 4, a confinement H- factor ≤3, a beta limit of approximately 2 times the first stability regime, a divertor heat flux of about 20 MW/m 2 , a Β max ≤ 18 T, and a TF coil average current density about 3 times that of ITER. The design constraints that bound the minimum COE are the allowable stresses in the TF coil, the neutral beam energy, and the 99% bootstrap current (essentially free current drive). 14 refs., 4 figs., 2 tabs

  1. Electricity

    International Nuclear Information System (INIS)

    Tombs, F.

    1983-01-01

    The subject is discussed, with particular reference to the electricity industry in the United Kingdom, under the headings; importance and scope of the industry's work; future fuel supplies (estimated indigenous fossil fuels reserves); outlook for UK energy supplies; problems of future generating capacity and fuel mix (energy policy; construction programme; economics and pricing; contribution of nuclear power - thermal and fast reactors; problems of conversion of oil-burning to coal-burning plant). (U.K.)

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

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

  3. Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

    2013-07-01

    In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

  4. Plasmachemical oxidation processes in a hybrid gas-liquid electrical discharge reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lukes, Petr; Locke, Bruce R [Department of Chemical and Biomedical Engineering, FAMU-FSU College of Engineering, Florida State University, 2525 Pottsdamer Street, Tallahassee, Florida (United States)

    2005-11-21

    Oxidation processes induced in water by pulsed electrical discharges generated simultaneously in the gas phase in close proximity to the water surface and directly in the liquid were investigated in a hybrid series gas-liquid electrical discharge reactor. The mechanism of phenol degradation was studied through its dependence on the gas phase and liquid phase compositions using pure argon and oxygen atmospheres above the liquid and different initial pH values in the aqueous solution. Phenol degradation was significantly enhanced in the hybrid-series reactor compared with the phenol removal by the single-liquid phase discharge reactor. Under an argon atmosphere the mechanism of phenol degradation was mainly caused by the electrophilic attack of OH{center_dot} radicals produced by the liquid phase discharge directly in water and OH{center_dot} radicals produced by the gas phase discharge at the gas-liquid interface. Under an oxygen atmosphere the formation of gaseous ozone dominated over the formation of OH{center_dot} radicals, and the contribution of the gas phase discharge in this case was determined mainly by the dissolution of gaseous ozone into the water and its subsequent interaction with phenol. At high pH phenol was degraded, in addition to the direct attack by ozone, also through indirect reactions of OH{center_dot} radicals formed via a peroxone process by the decomposition of dissolved ozone by hydrogen peroxide produced by the liquid phase discharge. Such a mechanism was proved by the detection of cis,cis-muconic acid and pH-dependent degradation of phenol, which resulted in significantly higher removal of phenol from alkaline solution observed under oxygen atmosphere than in argon.

  5. Plasmachemical oxidation processes in a hybrid gas-liquid electrical discharge reactor

    International Nuclear Information System (INIS)

    Lukes, Petr; Locke, Bruce R

    2005-01-01

    Oxidation processes induced in water by pulsed electrical discharges generated simultaneously in the gas phase in close proximity to the water surface and directly in the liquid were investigated in a hybrid series gas-liquid electrical discharge reactor. The mechanism of phenol degradation was studied through its dependence on the gas phase and liquid phase compositions using pure argon and oxygen atmospheres above the liquid and different initial pH values in the aqueous solution. Phenol degradation was significantly enhanced in the hybrid-series reactor compared with the phenol removal by the single-liquid phase discharge reactor. Under an argon atmosphere the mechanism of phenol degradation was mainly caused by the electrophilic attack of OH· radicals produced by the liquid phase discharge directly in water and OH· radicals produced by the gas phase discharge at the gas-liquid interface. Under an oxygen atmosphere the formation of gaseous ozone dominated over the formation of OH· radicals, and the contribution of the gas phase discharge in this case was determined mainly by the dissolution of gaseous ozone into the water and its subsequent interaction with phenol. At high pH phenol was degraded, in addition to the direct attack by ozone, also through indirect reactions of OH· radicals formed via a peroxone process by the decomposition of dissolved ozone by hydrogen peroxide produced by the liquid phase discharge. Such a mechanism was proved by the detection of cis,cis-muconic acid and pH-dependent degradation of phenol, which resulted in significantly higher removal of phenol from alkaline solution observed under oxygen atmosphere than in argon

  6. Power reactors in Member States. 1979 edition

    International Nuclear Information System (INIS)

    1979-01-01

    This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

  7. On generalization of electric field strength in longitudinally blown arcs

    OpenAIRE

    Yas'ko, O.I.; Esipchuk, A.M.; Qing, Z.; Schram, D.C.; Fauchais, P.

    1997-01-01

    Generalization of av. elec. field strength for different discharge conditions in longitudinally blown arcs is considered. Exptl. data for distinctive devices and different gases were used for phys. modeling. Anal. showed that heat transfer processes are responsible for I-E characteristic formation. Turbulent heat transfer is the most effective for atm. pressure discharges while convection plays the main role in vacuum arcs. A generalized I-E characteristic was obtained. [on SciFinder (R)

  8. Straight and chopped DC performance data for a General Electric 5BY436A1 DC shunt motor with a General Electric EV-1 controller

    Science.gov (United States)

    Edie, P. C.

    1981-01-01

    Both straight and chopped dc motor performance data for a General Electric 5BY436A1 motor with a General Electric EV-1 controller is presented in tabular and graphical formats. Effects of motor temperature and operating voltage are also shown. The maximum motor efficiency is approximately 85% at low operating temperatures in the straight dc mode. Chopper efficiency can be assumed to be 95% under all operating conditions. For equal speeds, the motor operated in the chopped mode develops slightly more torque and draws more current than it does in the straight mode.

  9. Modeling a clean energy standard for electricity: Policy design implications for emissions, supply, prices, and regions

    International Nuclear Information System (INIS)

    Paul, Anthony; Palmer, Karen; Woerman, Matt

    2013-01-01

    The electricity sector is responsible for roughly 40% of U.S. carbon dioxide (CO 2 ) emissions, and a reduction in CO 2 emissions from electricity generation is an important component of the U.S. strategy to reduce greenhouse gas emissions. Toward that goal, several proposals for a clean energy standard (CES) have been put forth, including one espoused by the Obama administration that calls for 80% clean electricity by 2035 phased in from current levels of roughly 40%. This paper looks at the effects of such a policy on CO 2 emissions from the electricity sector, the mix of technologies used to supply electricity, electricity prices, and regional flows of clean energy credits. The CES leads to a 30% reduction in cumulative CO 2 emissions between 2013 and 2035 and results in dramatic reductions in generation from conventional coal. The policy also results in fairly modest increases on national electricity prices, but this masks a wide variety of effects across regions. - Highlights: ► We model a clean energy standard (CES) for electricity at 80% by 2035. ► We analyze effects on CO 2 emissions, investment, prices, and credit trading. ► 80% CES leads to 30% reduction in cumulative CO 2 emissions by 2035. ► Modest national average electricity price increase masks regional heterogeneity

  10. Portland General Electric Company report on the operating and startup experience with control and instrumentation and electrical systems at the Trojan Nuclear Plant

    International Nuclear Information System (INIS)

    Zimmerman, G.A.

    1977-01-01

    The Trojan Nuclear Plant is an 1178 MWe nuclear plant located on the Columbia River 40 miles northwest of Portland, Oregon. The Nuclear Stream Supply System vendor is Westinghouse with a General Electric turbine generator. The reactor is rated and licensed for 3423 MWt (1178 MWe) and the turbine generator is designed for 3570 MWt(1219 MWe). The startup phase testing of Trojan commenced on November 21, 1975, upon receipt of our NRC Operating License. The startup testing program was completed on May 22, 1976, following 100 hours of full-power operation, at which time a scheduled summer maintenance outage began. Some of the highlights and milestones of the startup testing program are described

  11. 76 FR 66057 - North American Electric Reliability Corporation; Order Approving Regional Reliability Standard

    Science.gov (United States)

    2011-10-25

    ... Reliability Standard that is necessitated by a physical difference in the Bulk-Power System.\\7\\ \\7\\ Order No... Reliability Standards for the Bulk-Power System, Order No. 693, FERC Stats. & Regs. ] 31,242, order on reh'g... electric system event analyses and thereby improve system reliability by promoting improved system design...

  12. ELECTRICAL RESISTANCE HEATING OF SOILS AT C-REACTOR AT THE SAVANNAH RIVER SITE

    International Nuclear Information System (INIS)

    Blundy, R; Michael Morgenstern, M; Joseph Amari, J; Annamarie MacMurray, A; Mark Farrar, M; Terry Killeen, T

    2007-01-01

    Chlorinated solvent contamination of soils and groundwater is an endemic problem at the Savannah River Site (SRS), and originated as by-products from the nuclear materials manufacturing process. Five nuclear reactors at the SRS produced special nuclear materials for the nation's defense program throughout the cold war era. An important step in the process was thorough degreasing of the fuel and target assemblies prior to irradiation. Discharges from this degreasing process resulted in significant groundwater contamination that would continue well into the future unless a soil remediation action was performed. The largest reactor contamination plume originated from C-Reactor and an interim action was selected in 2004 to remove the residual trichloroethylene (TCE) source material by electrical resistance heating (ERH) technology. This would be followed by monitoring to determine the rate of decrease in concentration in the contaminant plume. Because of the existence of numerous chlorinated solvent sources around SRS, it was elected to generate in-house expertise in the design and operation of ERH, together with the construction of a portable ERH/SVE system that could be deployed at multiple locations around the site. This paper describes the waste unit characteristics, the ERH system design and operation, together with extensive data accumulated from the first deployment adjacent to the C-Reactor building. The installation heated the vadose zone down to 62 feet bgs over a 60 day period during the summer of 2006 and raised soil temperatures to over 200 F. A total of 730 lbs of trichloroethylene (TCE) were removed over this period, and subsequent sampling indicated a removal efficiency of 99.4%

  13. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

  15. A study on the implementation of the performance of electrical products standards

    Energy Technology Data Exchange (ETDEWEB)

    Berkowitz, M A

    1984-03-01

    The Steering Committee on Electrical Products (SCOPEP) has developed a number of performance standards for selected electrical products and product attributes, under the direction of the Canadian Standards Association. Although the development of these standards shows that a successful standards writing process is in place, the initiation and implementation has been achieved by one or more stakeholder groups spearheading the requirement for the standard. As a result, SCOPEP has been unable to achieve its full potential in its role of developing peformance standards for electrical products. A study was undertaken with the aim of developing a strategic plan for SCOPEP which would be supported by marketing and other strategies. The major elements of the strategy include: selecting those products for which the benefits of a standards program significantly outweigh the costs; agreeing on the use of a framework which rigorously evaluates the cost-benefits of developing performance standards; producing a critical mass of products through the selection process so that revenues from label fees will ultimately balance program costs and support a full time SCOPEP secretariat; aggressively marketing the standards in all sectors, using a balanced push-and-pull strategy; and developing a protagonist attitude so that standards programs implemented are credible to all SCOPEP stakeholders. Detailed cost-benefit tables are appended. 11 figs., 18 tabs.

  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. 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)

  18. 78 FR 4873 - Electrical Protective Equipment Standard and the Electric Power Generation, Transmission, and...

    Science.gov (United States)

    2013-01-23

    ... performing job duties. The training requirements of the Standard inform workers of the safety hazards of... this method, you must submit a copy of your comments and attachments to the OSHA Docket Office, Docket... Distribution Standard (Sec. 1910.269) Training Certification (Sec. 1910.269(a)(2)(vii)) This provision requires...

  19. The impact of electricity price changes on industrial prices and the general price level in Korea

    International Nuclear Information System (INIS)

    Lim, Seul-Ye; Yoo, Seung-Hoon

    2013-01-01

    Electricity has played an important role in the economic development of Korea and, thus, has become a critical factor in sustaining the well-being of the Korean people. This study attempts to investigate the impact of electricity price changes on industrial prices and the general price level using input–output (I–O) analysis. To this end, we apply the I–O price model to the 2011 I–O table recently produced by the Bank of Korea, paying particular attention to the electricity sector by considering it as exogenous and then investigating its impacts. The impacts of the electricity price changes on each industrial sector's prices and the general price level are quantitatively derived. For example, the overall impact of a 10% increase in electricity price on the Korean national economy is estimated to be 0.4367%. We also report the results from the model with the electricity sector endogenous and the model with endogenous electricity and labor sectors. This information can be usefully utilized in decision-making regarding price management for electricity. - Highlights: • We investigate the impact of electricity price changes on the Korean economy. • We use the input–output (I–O) analysis specifying the electricity sector as exogenous. • We apply the I–O price model to 2010 I–O table produced by the Bank of Korea. • The impact of a 10% increase in electricity price on the Korean economy is 0.2176%

  20. Halbach array motor/generators: A novel generalized electric machine

    Energy Technology Data Exchange (ETDEWEB)

    Merritt, B.T.; Post, R.F.; Dreifuerst, G.R.; Bender, D.A. [Lawrence Livermore National Lab., CA (United States)

    1995-02-01

    For many years Klaus Halbach has been investigating novel designs for permanent magnet arrays, using advanced analytical approaches and employing a keen insight into such systems. One of his motivations for this research was to find more efficient means for the utilization of permanent magnets for use in particle accelerators and in the control of particle beams. As a result of his pioneering work, high power free-electron laser systems, such as the ones built at the Lawrence Livermore Laboratory, became feasible, and his arrays have been incorporated into other particle-focusing systems of various types. This paper reports another, quite different, application of Klaus` work, in the design of high power, high efficiency, electric generators and motors. When tested, these motor/generator systems display some rather remarkable properties. Their success derives from the special properties which these arrays, which the authors choose to call {open_quotes}Halbach arrays,{close_quotes} possess.

  1. Clinch River Breeder Reactor Plant. License application, statement of general information

    International Nuclear Information System (INIS)

    1975-01-01

    Application is made for a reactor facility consisting of a liquid metal cooled reactor and steam generator system, a steam turbine driven electric generating system, electrical switchyard, and related auxiliaries and supporting structures. The primary system is located in an inert atmosphere in shielded vaults within a containment structure. Sodium coolant is used to remove heat from the core and radial blanket. Heat from the primary sodium is transferred in heat exchangers to non radioactive sodium which is used to convert feed-water into steam which is superheated to drive a tandem-compound generator. A single shaft multi-stage turbine generator produces 380 MW(e) with steam conditions of 1450 psig at 900 0 F. Fuel is sintered ceramic pellets of mixed uranium-plutonium oxides encapsulated in stainless steel. There are 198 fuel assemblies with each assembly consisting of 217 fuel rods placed in a hexagonal channel. Plutonium enrichment ranges from 1817 to 32.0 percent by weight. Axial blanket sections contain depleted UO 2 with 99.8 percent 238 U and 0.2 percent 235 U by weight. The proposed location of the plant is within the corporate limits of the city of Oak Ridge in Roane County, Tennessee. (U.S.)

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

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-10-01

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

  3. 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

  4. Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

    Science.gov (United States)

    Cook, William Gordon

    Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

  5. NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

    International Nuclear Information System (INIS)

    1992-08-01

    The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

  6. Managing realities of the new market uncertainty : standards development in the North American electricity industry

    International Nuclear Information System (INIS)

    Green, B.

    2003-01-01

    Standards for the electricity industry are developed to ensure quality and serve as a basis to which utilities should conform. Standards specify agreed upon properties for a manufactured product. They should be used for equipment specifications as well as operational procedures. Standardization is performed by regulators, transmission owners/operators, and organizations such as the National Electric Reliability Council (NERC), the Northeast Power Coordinating Council (NPCC), the North American Energy Standards Board (NAESB), and the Committee of Chief Risk Officers (CCRO). Before markets were opened to competition, operational standards were dictated by transmission owners and reliability issues were dealt with by NERC and NPCC. This presentation explained the process of standardization in the electric power industry in Canada, the derivation of standards, moving beyond NERC, and the transmission owners. Issues for Ontario Power Generation were highlighted. In contrast to the situation in the United States, there is no federal government backstop for developing Standards in Canada. There is no federal initiative toward open access. Canadian utilities participated in NERC, but compliance was voluntary. It is still questionable if Canadian utilities will implement NERC and NAESB Standards if they are codified

  7. Energy from sugarcane bagasse under electricity rationing in Brazil: a computable general equilibrium model

    International Nuclear Information System (INIS)

    Scaramucci, Jose A.; Perin, Clovis; Pulino, Petronio; Bordoni, Orlando F.J.G.; Cunha, Marcelo P. da; Cortez, Luis A.B.

    2006-01-01

    In the midst of the institutional reforms of the Brazilian electric sectors initiated in the 1990s, a serious electricity shortage crisis developed in 2001. As an alternative to blackout, the government instituted an emergency plan aimed at reducing electricity consumption. From June 2001 to February 2002, Brazilians were compelled to curtail electricity use by 20%. Since the late 1990s, but especially after the electricity crisis, energy policy in Brazil has been directed towards increasing thermoelectricity supply and promoting further gains in energy conservation. Two main issues are addressed here. Firstly, we estimate the economic impacts of constraining the supply of electric energy in Brazil. Secondly, we investigate the possible penetration of electricity generated from sugarcane bagasse. A computable general equilibrium (CGE) model is used. The traditional sector of electricity and the remainder of the economy are characterized by a stylized top-down representation as nested CES (constant elasticity of substitution) production functions. The electricity production from sugarcane bagasse is described through a bottom-up activity analysis, with a detailed representation of the required inputs based on engineering studies. The model constructed is used to study the effects of the electricity shortage in the preexisting sector through prices, production and income changes. It is shown that installing capacity to generate electricity surpluses by the sugarcane agroindustrial system could ease the economic impacts of an electric energy shortage crisis on the gross domestic product (GDP)

  8. Systems for controlling the electric power of a boiling water reactor power station

    International Nuclear Information System (INIS)

    Fukunishi, Koyu; Kiyokawa, Kazuhiro.

    1975-01-01

    Object: To achieve automatic increase and decrease of electric output in accordance with a predetermined rate of increase or decrease in output when the power output is raised or lowered. Structure: An electric output signal from an atomic power plant is led to a differentiating circuit through a smoothing circuit to produce a signal for rate of change of time, and an error signal between this signal and a preset signal produced from a circuit for a preset rate of change of output with time is supplied to an analog adjuster through a limiter. In this way, the flow rate in the reactor core is adjusted by a speed controller to obtain an output of a predetermined rate of increase. The difference signal between the electric output signal and a desired value signal is passed through an absolute circuit to a comparator circuit for comparison with a predetermined threshold value setting signal. The output signal of the comparator is used to operate a relay to open the contact so as to prevent an increase or decrease in the output beyond the required level. (Kamimura, M.)

  9. Review of operational experience with the gas-cooled Magnox reactors of the United Kingdom Central Electricity Generating Board

    International Nuclear Information System (INIS)

    Cave, L.; Clarke, A.W.

    1984-01-01

    The paper provides a review, which is mainly of a statistical nature, of 260 reactor years of operating experience which the (United Kingdom) Central Electricity Generating Board (CEGB) has obtained with its gas-cooled, graphite moderated Magnox reactors. The main emphasis in the review is on safety rather than on availability. Data are provided on the overall incidence and frequencies of faults and it is shown that the plant items which are predominantly responsible for recorded faults are the gas circulators and the turbo-alternators. Analysis of the reactor trip experience shows that the incidence of events which necessitate an automatic shutdown of the reactor has been about one per reactor year and that of other events leading to a reactor trip has not been much higher (1.4 per reactor year). As would be expected from the length of the operating experience, some relatively rare events have occurred (expected frequency 10 -2 per reactor year, or less) but on each occasion the reactor shutdown system and decay heat removal systems functioned satisfactorily. No overheating of, or damage to, the fuel occurred as a result of these rare events or of other, more frequent, faults. Analysis of the trend of failure rates has shown an improvement with time in nearly all safety-related items and external inspection of the primary coolant circuits has shown no significant deterioration with time. However, some derating of the reactors has been necessary to reduce the effects of oxidation of mild steel in CO 2 , in order to obtain optimum service lives. In spite of major differences between the systems, a comparison of the failure rates of analogous systems and plant items in PWRs and the Magnox reactors show a considerable similarity. Overall, the review of CEGB's operational experience with its Magnos reactors has shown that the frequencies of faults in systems and plant items has been satisfyingly low. (author)

  10. A Standardized Generalized Dimensionality Discrepancy Measure and a Standardized Model-Based Covariance for Dimensionality Assessment for Multidimensional Models

    Science.gov (United States)

    Levy, Roy; Xu, Yuning; Yel, Nedim; Svetina, Dubravka

    2015-01-01

    The standardized generalized dimensionality discrepancy measure and the standardized model-based covariance are introduced as tools to critique dimensionality assumptions in multidimensional item response models. These tools are grounded in a covariance theory perspective and associated connections between dimensionality and local independence.…

  11. Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2002-07-01

    The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

  12. LASIP-III, a generalized processor for standard interface files

    International Nuclear Information System (INIS)

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

    1976-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

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

  14. 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

  15. Mission operations for unmanned nuclear electric propulsion outer planet exploration with a thermionic reactor spacecraft.

    Science.gov (United States)

    Spera, R. J.; Prickett, W. Z.; Garate, J. A.; Firth, W. L.

    1971-01-01

    Mission operations are presented for comet rendezvous and outer planet exploration NEP spacecraft employing in-core thermionic reactors for electric power generation. The selected reference missions are the Comet Halley rendezvous and a Jupiter orbiter at 5.9 planet radii, the orbit of the moon Io. The characteristics of the baseline multi-mission NEP spacecraft are presented and its performance in other outer planet missions, such as Saturn and Uranus orbiters and a Neptune flyby, are discussed. Candidate mission operations are defined from spacecraft assembly to mission completion. Pre-launch operations are identified. Shuttle launch and subsequent injection to earth escape by the Centaur D-1T are discussed, as well as power plant startup and the heliocentric mission phases. The sequence and type of operations are basically identical for all missions investigated.

  16. The Perspective of Small and Medium Size Nuclear Power Reactors in the Brazilian Isolated Electricity Grids

    International Nuclear Information System (INIS)

    Moutinho dos Santos, E.

    1998-01-01

    As a consequence of the international crisis that has shaken the nuclear industry almost all over the world during the late 1980s, new academic contributions, which could settle a new basis for the nuclear energy in Brazil, have been scarce. This paper aims to partially fulfill this gap. We discuss some perspectives for the nuclear option to regain some prominence in the Brazilian energy matrix. Recent developments in the nuclear industry, including advances in plant design, have been opening interesting markets for small and medium nuclear power reactors (SMNPRs). We access the suitability and feasibility of such new technology in the Brazilian isolated electricity grids. We conclude by saying that, although the difficulties are still huge, SMNPRs may be a good strategy for Brazil to revitalize its nuclear policy. (author)

  17. ENERGY EFFICIENCY OF ELECTRIC HEATING OF REACTORS IN THE MANUFACTURE OF VARNISHES AND PAINTS

    Directory of Open Access Journals (Sweden)

    Tovajniansky L.L.

    2014-08-01

    Full Text Available The drawbacks of the traditional design of the heaters, which make known imperfections in manufacturing processes, realized with the use of electric heating. This determines the need for a radically new design of the heating devices. Created by high-temperature ceramics, characterized by abnormally high thermal stability and clarified the parameters that allow a certain degree change its thermal conductivity. On this basis the contact type ceramic heaters that provide thermal flow direction using different materials in the body of the heater - of high thermal conductivity, the surface facing the heat transfer and low which differs opposite sides of the heater are designed to eliminate the dissipation of heat into the surrounding space. This made it possible to equip the modern production paint industry energy efficient heating equipment with explosion and fire heating system reactors.

  18. Electric vehicles: The role and importance of standards in an emerging market

    International Nuclear Information System (INIS)

    Brown, Stephen; Pyke, David; Steenhof, Paul

    2010-01-01

    After nearly a century with the internal combustion engine dominating the personal transportation sector, it now appears that the electric vehicle is on the verge of experiencing rapid growth in both developed and developing vehicle markets. The broad-scale adoption of the electric vehicle could bring significant changes for society in terms of not only the technologies we use for personal transportation, but also moving our economies away from petroleum and lessoning the environmental footprint of transportation. This article investigates the role of standards, related training and certification for the electric vehicle. It is argued that the potential for the electric vehicle will be stunted without adequate attention being paid to standards, not only in terms of the speed of its uptake and smoothness of this transition, but also in terms of maintaining compatibility between jurisdictions, safety of the public, and helping to ensure environmental sustainability. We highlight a number of areas where new or adaptations of current standards, training and certification may be needed, notably in terms of batteries and charging infrastructures, electricity distribution and accounting for the environmental characteristics of this electricity, and different aspects of vehicle-to-grid and smart grid technologies.

  19. General Practice Messaging Standard Version 3.0, May 2014

    LENUS (Irish Health Repository)

    2018-01-25

    We addressed the research question "what factors enable the successful development and implementation of a frail older person\\'s pathway within the acute setting". A rapid realist review (RRR) was conducted by adopting the RAMESES standards. We began with a sample of 232 articles via database searches supplemented with 94 additional records including inputs from a twitter chat and a hospital site visit. Our final sample consisted of 18 documents. Following review and consensus by an expert panel we identified a conceptual model of context-mechanism-(resources)-outcomes. There was overall agreement frailty should be identified at the front door of the acute hospital. Significant challenges identified related to organisational boundaries both within the acute setting and externally, the need to shift outcomes to patient orientated ones, to support staff to sustain the pathway by providing ongoing education and by providing role clarity. RRRs can support research such as the systematic approach to improving care for frail older adults (SAFE) study by producing accounts of what works based on a wide range of sources and innovative engagement with stakeholders. It is evident from our provisional model that numerous factors need to combine and interact to enable and sustain a successful frail older person\\'s pathway.

  20. An International Survey of Electric Storage Tank Water Heater Efficiency and Standards

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Alissa; Lutz, James; McNeil, Michael A.; Covary, Theo

    2013-11-13

    Water heating is a main consumer of energy in households, especially in temperate and cold climates. In South Africa, where hot water is typically provided by electric resistance storage tank water heaters (geysers), water heating energy consumption exceeds cooking, refrigeration, and lighting to be the most consumptive single electric appliance in the home. A recent analysis for the Department of Trade and Industry (DTI) performed by the authors estimated that standing losses from electric geysers contributed over 1,000 kWh to the annual electricity bill for South African households that used them. In order to reduce this burden, the South African government is currently pursuing a programme of Energy Efficiency Standards and Labelling (EES&L) for electric appliances, including geysers. In addition, Eskom has a history of promoting heat pump water heaters (HPWH) through incentive programs, which can further reduce energy consumption. This paper provides a survey of international electric storage water heater test procedures and efficiency metrics which can serve as a reference for comparison with proposed geyser standards and ratings in South Africa. Additionally it provides a sample of efficiency technologies employed to improve the efficiency of electric storage water heaters, and outlines programs to promote adoption of improved efficiency. Finally, it surveys current programs used to promote HPWH and considers the potential for this technology to address peak demand more effectively than reduction of standby losses alone

  1. Basic guidelines to introduce electric circuit simulation software in a general physics course

    Science.gov (United States)

    Moya, A. A.

    2018-05-01

    The introduction of electric circuit simulation software for undergraduate students in a general physics course is proposed in order to contribute to the constructive learning of electric circuit theory. This work focuses on the lab exercises based on dc, transient and ac analysis in electric circuits found in introductory physics courses, and shows how students can use the simulation software to do simple activities associated with a lab exercise itself and with related topics. By introducing electric circuit simulation programs in a general physics course as a brief activitiy complementing lab exercise, students develop basic skills in using simulation software, improve their knowledge on the topology of electric circuits and perceive that the technology contributes to their learning, all without reducing the time spent on the actual content of the course.

  2. 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.)

  3. Status of development and licensing support for advanced liquid metal reactors in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, D R [Argonne National Laboratory, Argonne, IL (United States); Gyorey, G [General Electric, San Jose, CA (United States)

    1991-07-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  4. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment

  5. Safety aspects of the US advanced LMR [liquid metal reactor] design

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S.

    1989-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs

  6. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  7. 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...

  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. Potential of Electric Power Production from Microbial Fuel Cell (MFC) in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

    Science.gov (United States)

    Zaman, Badrus; Wardhana, Irawan Wisnu

    2018-02-01

    Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media). Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day) operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

  10. Potential of Electric Power Production from Microbial Fuel Cell (MFC in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

    Directory of Open Access Journals (Sweden)

    Zaman Badrus

    2018-01-01

    Full Text Available Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media. Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

  11. 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.)

  12. 18 CFR 38.2 - Incorporation by reference of North American Energy Standards Board Wholesale Electric Quadrant...

    Science.gov (United States)

    2010-04-01

    ... reference of North American Energy Standards Board Wholesale Electric Quadrant standards. 38.2 Section 38.2 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY... UTILITIES § 38.2 Incorporation by reference of North American Energy Standards Board Wholesale Electric...

  13. Development of a general learning algorithm with applications in nuclear reactor systems

    International Nuclear Information System (INIS)

    Brittain, C.R.; Otaduy, P.J.; Perez, R.B.

    1989-12-01

    The objective of this study was development of a generalized learning algorithm that can learn to predict a particular feature of a process by observation of a set of representative input examples. The algorithm uses pattern matching and statistical analysis techniques to find a functional relationship between descriptive attributes of the input examples and the feature to be predicted. The algorithm was tested by applying it to a set of examples consisting of performance descriptions for 277 fuel cycles of Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR). The program learned to predict the critical rod position for the HFIR from core configuration data prior to reactor startup. The functional relationship bases its predictions on initial core reactivity, the number of certain targets placed in the center of the reactor, and the total exposure of the control plates. Twelve characteristic fuel cycle clusters were identified. Nine fuel cycles were diagnosed as having noisy data, and one could not be predicted by the functional relationship. 13 refs., 6 figs

  14. Development of a general learning algorithm with applications in nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Brittain, C.R.; Otaduy, P.J.; Perez, R.B.

    1989-12-01

    The objective of this study was development of a generalized learning algorithm that can learn to predict a particular feature of a process by observation of a set of representative input examples. The algorithm uses pattern matching and statistical analysis techniques to find a functional relationship between descriptive attributes of the input examples and the feature to be predicted. The algorithm was tested by applying it to a set of examples consisting of performance descriptions for 277 fuel cycles of Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR). The program learned to predict the critical rod position for the HFIR from core configuration data prior to reactor startup. The functional relationship bases its predictions on initial core reactivity, the number of certain targets placed in the center of the reactor, and the total exposure of the control plates. Twelve characteristic fuel cycle clusters were identified. Nine fuel cycles were diagnosed as having noisy data, and one could not be predicted by the functional relationship. 13 refs., 6 figs.

  15. Guide to General Atomic studies of hypothetical nuclear driven accidents for the Fort St. Vrain reactor

    International Nuclear Information System (INIS)

    Wei, T.; Tobias, M.

    1974-03-01

    The work of the General Atomic Company (GAC) in preparing those portions of the Final Safety Analysis Report for the Fort St. Vrain Reactor (FSV) having to do with hypothetical nuclear driven accidents has been reviewed and a guide to this literature has been prepared. The sources for this study are the Final Safety Analysis Report itself, the Quarterly and Monthly Progress Reports, Topical Reports, and Technical Specifications. The problems considered and the methods used are outlined. An appendix gives a systematic analysis which was used as a guide in organizing the references. (U.S.)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  17. The contract of design of an atomic reactor system in Yonggwang - 5, 6 nuclear power plant

    International Nuclear Information System (INIS)

    1995-05-01

    This is a contract of design of an atomic reactor system in Yonggwang 5, 6 nuclear power plant. It has the general contract condition. In the appendix, it indicates the detail regulations between two parties which are the coverage of division on the responsibility, schedule of the delivery, standard of the technology, guarantee, drawing and paper support of the Korea Electric Power Corporation, support of technology drill, test, regulations of code and standard and list of items and prices.

  18. Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    Nero, A.V.; Farnaam, M.R.K.

    1977-01-01

    This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

  19. A Deeper Glimpse into the National Core Arts Standards for General Music

    Science.gov (United States)

    Zaffini, Erin Dineen

    2018-01-01

    The National Core Arts Standards in general music provide some exciting possibilities for music growth and understanding among our students. For those of us who are still unsure of how to read the standards or implement them in our classrooms, the standards also present some challenges for music educators. This article provides a deeper look into…

  20. 2016 Standard Scenarios Report: A U.S. Electricity Sector Outlook

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Wesley [National Renewable Energy Lab. (NREL), Golden, CO (United States); Mai, Trieu [National Renewable Energy Lab. (NREL), Golden, CO (United States); Logan, Jeffrey [National Renewable Energy Lab. (NREL), Golden, CO (United States); Steinberg, Daniel [National Renewable Energy Lab. (NREL), Golden, CO (United States); McCall, James [National Renewable Energy Lab. (NREL), Golden, CO (United States); Richards, James [National Renewable Energy Lab. (NREL), Golden, CO (United States); Sigrin, Benjamin [National Renewable Energy Lab. (NREL), Golden, CO (United States); Porro, Gian [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-11-01

    The National Renewable Energy Laboratory is conducting a study sponsored by the Office of Energy Efficiency and Renewable Energy (EERE) that aims to document and implement an annual process designed to identify a realistic and timely set of input assumptions (e.g., technology cost and performance, fuel costs), and a diverse set of potential futures (standard scenarios), initially for electric sector analysis.

  1. Physical and electrical bandwidths of integrated photodiodes in standard CMOS technology

    NARCIS (Netherlands)

    Radovanovic, S.; Annema, Anne J.; Nauta, Bram

    2003-01-01

    The influence of different geometries (layouts) and structures of high-speed photodiodes in fully standard 0.18 μm CMOS technology on their intrinsic (physical) and electrical bandwidths is analyzed. Three photodiode structures are studied: nwell/p-substrate, p+/nwell/p-substrate and p+/nwell. The

  2. 2017 Standard Scenarios Report: A U.S. Electricity Sector Outlook

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Wesley J. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Mai, Trieu T. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Richards, James [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Das, Paritosh [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Donohoo-Vallett, Paul [US Department of Energy, Washington, DC (United States). Office of Energy Efficiency and Renewable Energy (EERE)

    2017-10-03

    The 2017 Standard Scenarios includes a suite of U.S. electricity sector scenarios. The report explores four power sector storylines, including the growth in natural gas and renewable energy, the relative competitiveness of wind and solar PV, the potential impact of low-cost battery storage, and the impact of nuclear lifetimes on the capacity expansion of the power sector.

  3. The standard model prediction for the electric dipole moment of the electron

    International Nuclear Information System (INIS)

    Hoogeveen, F.

    1990-01-01

    The electric dipole moment of the electron is calculated within the standard model with three generations of quarks. Depending on the values of some unknown parameters like the top quark mass and the CP-violating phase δ in the Kobayashi-Maskawa matrix, its value is of the order of magnitude of 2x10 -38 vertical strokeevertical stroke cm. (orig.)

  4. Retail electricity price savings from compliance flexibility in GHG standards for stationary sources

    International Nuclear Information System (INIS)

    Burtraw, Dallas; Woerman, Matt; Paul, Anthony

    2012-01-01

    The EPA will issue rules regulating greenhouse gas (GHG) emissions from existing steam boilers and refineries in 2012. A crucial issue affecting the scope and cost of emissions reductions will be the potential introduction of flexibility in compliance, including averaging across groups of facilities. This research investigates the role of compliance flexibility for the most important of these source categories—existing coal-fired power plants—that currently account for one-third of national emissions of carbon dioxide, the most important greenhouse gas. We find a flexible standard, calibrated to achieve the same emissions reductions as a traditional(inflexible) approach, reduces the increase in electricity price by 60 percent and overall costs by two-thirds in 2020. The flexible standard also leads to substantially more investment to improve the operating efficiency of existing facilities, whereas the traditional standard leads to substantially greater retirement of existing facilities. - Highlights: ► The U.S. EPA will regulate GHG emissions from power plants under the Clean Air Act. ► We compare a flexible standard with fleet-wide averaging to a traditional standard. ► Flexible standard reduces the increase in electricity price by 60percent in 2020. ► Flexible standard reduces the increase in overall costs by two-thirds in 2020. ► Flexible standard leads to more efficiency investment and less capacity retirement.

  5. Harmonization of IEEE323 and IEC60780 standards For Environmental Qualificaiton of Electric Equipment

    International Nuclear Information System (INIS)

    Kim, Jong Seog

    2009-01-01

    IEEE323 standard has been widely used for the qualification of electric equipment in Asian pacific area while IEC6070 has been mostly used in European area. Since each plant use different standard for environmental qualification, manufacturer has to perform the qualification test twice in accordance with each standard. Problem also can be happened in the plant site when they are going to purchase equipment qualified by different qualification standard which are not used in his plant. The need of harmonization of each standard has been raised several years and it is known that some studies are in progress by IEEE committee. KEPRI has a plan of comparing EQ relative standards of IEEE, IEC and RCC in 2009. In this paper, brief comparing result between IEEE323 and IEC60780 and the proper harmonization method is introduced

  6. High temperature blankets for non-electrical/electrical applications of fusion reactors: Progress report, July 15, 1983--November 30, 1984

    International Nuclear Information System (INIS)

    Ribe, F.L.; Woodruff, G.L.

    1988-01-01

    We report a continuation of work done in collaboration with the Lawrence Livermore National Laboratory (LLNL) on design studies of the tandem-mirror fusion reactor (TMR) coupled to the General Atomic (GA) sulfur-iodine thermochemical process for producing hydrogen. During this report period the emphasis was on a solid-breeder gas cooled ''cannister'' blanket for TMR-based hydrogen production. This work was integrated with the Department of Energy (DOE), Office of Fusion Energy (OFE) Blanket Comparison and Selection Study, coordinated by the Argonne National Laboratory (ANL). The areas investigated by the two principal investigators and their students were the following: Plasma engineering of the TMR, including the magnets. Neutronics transport support for the synfuel blanket and shield. Completion of studies of the GA sulfur-iodine process. Under subcontract D.S. Rowe of Rowe and Associates worked with both UW and LLNL personnel on Mechanical design and thermal hydraulics of a high temperature, solid breeder blanket. 2 refs., 3 figs

  7. 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)

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

    International Nuclear Information System (INIS)

    Keutner, H.

    1979-01-01

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

  9. IEEE Std 317-1972: IEEE standard for electric penetration assemblies in containment structures for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This standard prescribes the mechanical, electrical, and test requirements for the design, construction, and installation of electric penetration assemblies in containment structures for stationary nuclear power generating stations. The electric conductor and insulation characteristics of external circuits which connect to penetration assemblies are beyond the scope of these criteria. If there should be any conflict between this standard and those documents referenced herein, this standard shall take precedence over the referenced documents

  10. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  11. Technical evaluation report on the monitoring of electric power to the reactor-protection system for the Brunswick Steam Electric Plant, Units 1 and 2

    International Nuclear Information System (INIS)

    Selan, J.C.

    1982-01-01

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Brunswick Steam Electric Plant, Units 1 and 2. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications with time delays verified by GE, will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

  12. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  13. General solution for calculating polarization electric fields in the auroral ionosphere and application examples

    Science.gov (United States)

    Amm, O.; Fujii, R.; VanhamäKi, H.; Yoshikawa, A.; Ieda, A.

    2013-05-01

    We devise an approach to calculate the polarization electric field in the ionosphere, when the ionospheric conductances, the primary (modeled) or the total (measured) electric field, and the Cowling efficiency are given. In contrast to previous studies, our approach is a general solution which is not limited to specific geometrical setups, and all parameters may have any kind of spatial dependence. The solution technique is based on spherical elementary current (vector) systems (SECS). This way, we avoid the need to specify explicit boundary conditions for the searched polarization electric field of its potential which would be required if the problem was solved in a differential equation approach. Instead, we solve an algebraic matrix equation, and the implicit boundary condition that the divergence of the polarization electric field vanishes outside our analysis area is sufficient. In order to illustrate our theory, we then apply it to two simple models of auroral electrodynamic situations, the first being a mesoscale strong conductance enhancement in the early morning sector within a relatively weak southward primary electric field, and a morning sector auroral arc with only a weak conductance enhancement, but a large southward primary electric field at the poleward flank of the arc. While the significance of the polarization electric field for maximum Cowling efficiency is large for the first case, it is rather minor for the second one. Both models show that the polarization electric field effect may not only change the magnitude of the current systems but also their overall geometry. Furthermore, the polarization electric field may extend into regions where the primary electric field is small, thus even dominating the total electric field in these regions. For the first model case, the total Joule heating integrated over the analysis area decreases by a factor of about 4 for maximum Cowling efficiency as compared to the case of vanishing Cowling efficiency

  14. General principles of nuclear safety management related to research reactor decommissioning

    International Nuclear Information System (INIS)

    Banciu, Ortenzia; Vladescu, Gabriela

    2003-01-01

    The paper contents the general principles applicable to the decommissioning of research reactors to ensure a proper nuclear safety management, during both decommissioning activities and post decommissioning period. The main objective of decommissioning is to ensure the protection of workers, population and environment against all radiological and non-radiological hazards that could result after a reactor shutdown and dismantling. In the same time, it is necessary, by some proper provisions, to limit the effect of decommissioning for the future generation, according to the new Romanian, IAEA and EU Norms and Regulations. Assurance of nuclear safety during decommissioning process involves, in the first step, to establish of some safety principles and requirements to be taken into account during whole process. In the same time, it is necessary to perform a series of analyses to ensure that the whole process is conducted in a planned and safe manner. The general principles proposed for a proper management of safety during research reactor decommissioning are as follows: - Set-up of all operations included in a Decommissioning Plan; - Set-up and qualitative evaluation of safety problems, which could appear during normal decommissioning process, both radiological and nonradiological risks for workers and public; - Set-up of accident list related to decommissioning process the events that could appear both due to some abnormal working conditions and to some on-site and off-site events like fires, explosions, flooding, earthquake, etc.); - Development and qualitative/ quantitative evaluation of scenarios for each incidents; - Development (and evaluation) of safety indicator system. The safety indicators are the most important tools used to assess the level of nuclear safety during decommissioning process, to discover the weak points and to establish safety measures. The paper contains also, a safety case evaluation (description of facility according to the decommissioning

  15. A Step Towards Electric Propulsion Testing Standards: Pressure Measurements and Effective Pumping Speeds

    Science.gov (United States)

    Dankanich, John W.; Swiatek, Michael W.; Yim, John T.

    2012-01-01

    The electric propulsion community has been implored to establish and implement a set of universally applicable test standards during the research, development, and qualification of electric propulsion systems. Existing practices are fallible and result in testing variations which leads to suspicious results, large margins in application, or aversion to mission infusion. Performance measurements and life testing under appropriate conditions can be costly and lengthy. Measurement practices must be consistent, accurate, and repeatable. Additionally, the measurements must be universally transportable across facilities throughout the development, qualification, spacecraft integration and on-orbit performance. A preliminary step to progress towards universally applicable testing standards is outlined for facility pressure measurements and effective pumping speed calculations. The standard has been applied to multiple facilities at the NASA Glenn Research Center. Test results and analyses of universality of measurements are presented herein.

  16. Electrochemical Recovery of Gold from Waste Electric and Electronic Equipment Using Circulating Particulate Bed Reactor (CPBE)

    Science.gov (United States)

    Ravinder, T.; Ali, U. F. M.; Ridwan, F. M.; Ibrahim, N.; Azmi, N. H.

    2017-06-01

    The utilization of electrochemical process recovery involving low reactant concentrations of metal requires electrodes with high mass transport rates and specific surface areas. This is essential to increase cross-sectional current densities whilst optimizing the capital and operating costs. Experimental results demonstrated that Circulating Particulate Bed Reactor (CPBE) is suitable for the recovery of low concentrations of gold from aqueous chloride solution containing {{AuCl}}4- and {{AuCl}}2- of less than 0.5 mol m-3(< 102 g m-3). Elemental gold was successfully obtained on 0.5-1 mm gr particles in an electrochemical reactor incorporating a cation- permeable membrane and operated in bath recycle mode. Depletion to concentration < 5 × 10-3 mol m-3 (< 1 g m-3) appeared to be mass transport controlled at an applied potential of +0.20 V (SCE), specific electrical energy consumption (SEEC) of ca. 800-1300 kWh h (tonne Au)-1 for cell voltages (U) of 2.0-3.0 V, and fractional current efficiencies of ca. 0.95. However, atomic absorption and UV spectrophotometry established that as the ([{{AuCl}}4-+[{{AuCl}}2-]) concentration decayed, the [{{AuCl}}4-]:[{{AuCl}}2-] molar ratio changed. A multi-step mechanism for reduction of {{AuCl}}4- ions explained this behavior in terms of changing overpotentials for {{AuCl}}4- and {{AuCl}}2- reduction as total dissolved gold concentration decreased. In addition, SEM images confirmed that adherent and coherent Au deposits were achieved with CPBE for Au deposition under mass transport control at 0.20 V (SCE).

  17. Validation and standardization of the Generalized Anxiety Disorder Screener (GAD-7) in the general population.

    Science.gov (United States)

    Löwe, Bernd; Decker, Oliver; Müller, Stefanie; Brähler, Elmar; Schellberg, Dieter; Herzog, Wolfgang; Herzberg, Philipp Yorck

    2008-03-01

    The 7-item Generalized Anxiety Disorder Scale (GAD-7) is a practical self-report anxiety questionnaire that proved valid in primary care. However, the GAD-7 was not yet validated in the general population and thus far, normative data are not available. To investigate reliability, construct validity, and factorial validity of the GAD-7 in the general population and to generate normative data. Nationally representative face-to-face household survey conducted in Germany between May 5 and June 8, 2006. Five thousand thirty subjects (53.6% female) with a mean age (SD) of 48.4 (18.0) years. The survey questionnaire included the GAD-7, the 2-item depression module from the Patient Health Questionnaire (PHQ-2), the Rosenberg Self-Esteem Scale, and demographic characteristics. Confirmatory factor analyses substantiated the 1-dimensional structure of the GAD-7 and its factorial invariance for gender and age. Internal consistency was identical across all subgroups (alpha = 0.89). Intercorrelations with the PHQ-2 and the Rosenberg Self-Esteem Scale were r = 0.64 (P < 0.001) and r = -0.43 (P < 0.001), respectively. As expected, women had significantly higher mean (SD) GAD-7 anxiety scores compared with men [3.2 (3.5) vs. 2.7 (3.2); P < 0.001]. Normative data for the GAD-7 were generated for both genders and different age levels. Approximately 5% of subjects had GAD-7 scores of 10 or greater, and 1% had GAD-7 scores of 15 or greater. Evidence supports reliability and validity of the GAD-7 as a measure of anxiety in the general population. The normative data provided in this study can be used to compare a subject's GAD-7 score with those determined from a general population reference group.

  18. Electrical Capacitance Volume Tomography for the Packed Bed Reactor ISS Flight Experiment

    Science.gov (United States)

    Marashdeh, Qussai; Motil, Brian; Wang, Aining; Liang-Shih, Fan

    2013-01-01

    Fixed packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a highly desirable unit operation for long duration life support systems in space. NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. To validate these models, the instantaneous distribution of the gas and liquid phases must be measured.Electrical Capacitance Volume Tomography (ECVT) is a non-invasive imaging technology recently developed for multi-phase flow applications. It is based on distributing flexible capacitance plates on the peripheral of a flow column and collecting real-time measurements of inter-electrode capacitances. Capacitance measurements here are directly related to dielectric constant distribution, a physical property that is also related to material distribution in the imaging domain. Reconstruction algorithms are employed to map volume images of dielectric distribution in the imaging domain, which is in turn related to phase distribution. ECVT is suitable for imaging interacting materials of different dielectric constants, typical in multi-phase flow systems. ECVT is being used extensively for measuring flow variables in various gas-liquid and gas-solid flow systems. Recent application of ECVT include flows in risers and exit regions of circulating fluidized beds, gas-liquid and gas-solid bubble columns, trickle beds, and slurry bubble columns. ECVT is also used to validate flow models and CFD simulations. The technology is uniquely qualified for imaging phase concentrations in packed bed reactors for the ISS flight experiments as it exhibits favorable features of compact size, low profile sensors, high imaging speed, and

  19. The development of professional practice standards for Australian general practice nurses.

    Science.gov (United States)

    Halcomb, Elizabeth; Stephens, Moira; Bryce, Julianne; Foley, Elizabeth; Ashley, Christine

    2017-08-01

    The aim of this study was to explore the current role of general practice nurses and the scope of nursing practice to inform the development of national professional practice standards for Australian general practice nurses. Increasing numbers of nurses have been employed in Australian general practice to meet the growing demand for primary care services. This has brought significant changes to the nursing role. Competency standards for nurses working in general practice were first developed in Australia in 2005, but limited attention has been placed on articulating the contemporary scope of practice for nurses in this setting. Concurrent mixed methods design. Data collection was conducted during 2013-2014 and involved two online surveys of Registered and Enrolled Nurses currently working in general practice, a series of 14 focus groups across Australia and a series of consultations with key experts. Data collection enabled the development of 22 Practice Standards separated into four domains: (i) Professional Practice; (ii) Nursing Care; (iii) General Practice Environment and (iv) Collaborative Practice. To differentiate the variations in enacting these Standards, performance indicators for the Enrolled Nurse, Registered Nurse and Registered Nurse Advanced Practice are provided under each Standard. The development of national professional practice standards for nurses working in Australian general practice will support ongoing workforce development. These Standards are also an important means of articulating the role and scope of the nurses' practice for both consumers and other health professionals, as well as being a guide for curriculum development and measurement of performance. © 2017 John Wiley & Sons Ltd.

  20. The HTR-10 test reactor project and potential use of HTGR for non-electric application in China

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Xu Yuanhui; Wu Zhongxin

    1997-01-01

    Coal is the dominant source of energy in China. This use of coal results in two significant problems for China; it is a major burden on the train, road and waterway transportation infrastructures and it is a significant source of environmental pollution. In order to ease the problems caused by the burning of coal and to help reduce the energy supply shortage in China, national policy has directed the development of nuclear power. This includes the erection of nuclear power plants with water cooled reactors and the development of advanced nuclear reactor types, specifically, the high temperature gas cooled reactor (HTGR). The HTGR was chosen for its favorable safety features and its ability to provide high reactor outlet coolant temperatures for efficient power generation and high quality process heat for industrial applications. As the initial modular HTGR development activity within the Chinese High Technology Programme, a 10MW helium cooled test reactor is currently under construction on the site of the Institute of Nuclear Energy Technology northwest of Beijing. This plant features a pebble-bed helium cooled reactor with initial criticality anticipated in 1999. There will be two phases of high temperature heat utilization from the HTR-10. The first phase will utilize a reactor outlet temperature of 700 deg. C with a steam generator providing steam for a steam turbine cycle which works on an electrical/heat co-generation basis. The second phase is planned for a core outlet temperature of 900 deg. C to investigate a steam cycle/gas turbine combined cycle system with the gas turbine and the steam cycle being independently parallel in the secondary side of the plant. This paper provides a review of the technical design, licensing, safety and construction schedule for the HTR-10. It also addresses the potential uses of the HTGR for non-electric applications in China including process steam for the petrochemical industry, heavy oil recovery, coal conversion and

  1. Electricity savings from implementation of minimum energy efficiency standard for TVs in Malaysia

    Energy Technology Data Exchange (ETDEWEB)

    Varman, M.; Masjuki, H.H.; Mahlia, T.M.I. [University of Malaya, Kuala Lumpur (Malaysia). Department of Mechanical Engineering

    2005-06-01

    The popularization of 24 h pay-TV, interactive video games, web-TV, VCD and DVD in Malaysia are poised to have a large impact on overall TV electricity consumption in the country. With the increasing of overall TV energy consumption, energy efficiency standards are one of highly effective policies for decreasing electricity consumption in the residential sector. Energy efficiency standards are also capable of reducing consumer's electricity bill and contribute towards positive environmental impacts. This paper attempts to predict the amount of energy that can be saved in the residential sector by implementing minimum energy efficiency standard for television sets in Malaysia. Over the past 30 years, television ownership in Malaysian residents has increased from 186,036 units in 1970 to 2,741,640 units in 1991. This figure is expected to reach 6,201,316 units in the year 2010. Hence, efficiency improvement for this appliance will have a significant impact on the future of electricity consumption in this country. (author)

  2. Report on the results of the FY1999 standardization for new standard interface protocol for electrical measuring use; 1999 nendo shinki sangyo ikusei sokkogata kokusai hyojun kaihatsu jigyo seika hokokusho. Denki keisokukiyo shinhyojun interface purotokoru no hyojunka

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    The purpose of this project is to create a new protocol for electric measuring instruments that use a built-in the Universal Serial Bus (USB) interface which is no incorporated in PCs. This new interface will replace the existing GPIB interface (IEEE-488), which is currently used in electric measuring instruments as standard interface. Our goal is to create a protocol that will provide the same data rate as GPIB and better connectivity. To evaluate our new protocol, we create virtual instruments using evaluation boards for USB chips, as well as software, and performed transfer tests to isolate and solve problems found in the tests. We will make the protocol and the software available for general use, and register the protocol as a standard to the USB Developer's Forum, a body that manages USB standards, to make it as a de facto standard. Ultimately we are aiming at making it an international standard. (NEDO)

  3. IEEE [Institute of Electrical and Electronics Engineers] standards and nuclear software quality engineering

    International Nuclear Information System (INIS)

    Daughtrey, T.

    1988-01-01

    Significant new nuclear-specific software standards have recently been adopted under the sponsorship of the American Nuclear Society and the American Society of Mechanical Engineers. The interest of the US Nuclear Regulatory Commission has also been expressed through their issuance of NUREG/CR-4640. These efforts all indicate a growing awareness of the need for thorough, referenceable expressions of the way to build in and evaluate quality in nuclear software. A broader professional perspective can be seen in the growing number of software engineering standards sponsored by the Institute of Electrical and Electronics Engineers (IEEE) Computer Society. This family of standards represents a systematic effort to capture professional consensus on quality practices throughout the software development life cycle. The only omission-the implementation phase-is treated by accepted American National Standards Institute or de facto standards for programming languages

  4. Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2015-01-01

    The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

  5. Environmental assessment for the electric and hybrid vehicle demonstration project, performance standards and financial incentives

    Energy Technology Data Exchange (ETDEWEB)

    LaBelle, S. J.

    1978-10-01

    The assessment is concerned with the impacts of the demonstration of electric and hybrid vehicles acquired to fulfill certain requirements of the Electric and Hybrid Vehicle Research, Development, and Demonstration Act, PL 94-413 as amended. The financial incentives programs and vehicle performance standards associated with the demonstration are also covered. Not included is an assessment of the long term effects of EHV commercialization and of the research and development program being carried out simultaneously with the demonstration, also in response to PL 94-413. These federal actions will be included in a programmatic environmental assessment scheduled for completion in FY 79.

  6. Etalon (standard) for surface potential distribution produced by electric activity of the heart.

    Science.gov (United States)

    Szathmáry, V; Ruttkay-Nedecký, I

    1981-01-01

    The authors submit etalon (standard) equipotential maps as an aid in the evaluation of maps of surface potential distributions in living subjects. They were obtained by measuring potentials on the surface of an electrolytic tank shaped like the thorax. The individual etalon maps were determined in such a way that the parameters of the physical dipole forming the source of the electric field in the tank corresponded to the mean vectorcardiographic parameters measured in a healthy population sample. The technique also allows a quantitative estimate of the degree of non-dipolarity of the heart as the source of the electric field.

  7. Assessment of two small-sized innovative nuclear reactors for electricity generation in Brazil using INPRO methodology

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho; Sefidvash, Farhang

    2009-01-01

    This paper presents the main results of the assessment study of two small-sized innovative reactors for electricity generation in Brazil using the methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO was initiated in 2001 and has the main objective of helping to ensure that nuclear energy is available to contribute in a sustainable manner to the energy needs of the 21st century. Brazil joined the INPRO project since its beginning and in 2005 submitted a proposal for the assessment using INPRO methodology of two small-sized reactors (IRIS - International Reactor Innovative and Secure, and FBNR - Fixed Bed Nuclear Reactor) as potential components of an innovative nuclear energy system (INS) completed by a conventional open nuclear fuel cycle based on enriched uranium. The scope of this assessment study was restricted to the reactor component of the INS and to the methodology areas of economics and safety for IRIS, and proliferation resistance and safety for FBNR. The results indicate that both IRIS and FBNR innovative designs comply mostly with the basic principles of the areas assessed and have potential to comply with the remaining ones. (author)

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

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

  9. General views about specimen irradiations in reactors; Considerations generales sur'les irradiations d'echantillons dans les reacteurs

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    Specimen irradiation of fissile or non-fissile materials, carried out under circumstances becoming more and more severe and in reactor of increasing flux bas led to an evolution of irradiation rigs. A survey of the problems arising from irradiating under these various circumstances leads to conclude that it is possible to devise one capsule type suitable to every particular case, and that in a wide temperature range. Consequently, once the various irradiation-parameters known, a general method of calculation can be followed so as to determine the various sizes of the parts constituting the capsule. These theoretical calculations might sometimes be corrected through benefits gained from previous irradiations. Similarly, practical experimentation might allow to foresee more handy assembling of the capsule, specimen loading-and unloading being easier at the same time. (author) [French] L'irradiation d'echantillons, fissiles ou non fissiles, dans des conditions imposees de plus en plus strictes et dans des reacteurs a flux de plus en plus eleve, a eu pour consequence une evolution dans la conception des dispositifs d'irradiation. Lorsqu'on examine les problemes souleves par ces differentes irradiations, on en conclut qu'il est possible de concevoir un type de capsule capable de donner satisfaction dans chaque cas particulier, et ce, dans une tres large gamme de temperature. Par consequent, les differents parametres de l'irradiation etant connus, une methode generale de calcul peut etre suivie pour determiner les differentes cotes des pieces constitutives de la capsule. Ces calculs theoriques devront quelquefois etre corriges grace aux enseignements tires d'irradiations precedentes. De meme, l'experience acquise permettra d'envisager un montage plus aise de la capsule, tout en facilitant l'enfournement et le defournement des echantillons.

  10. Results from the Operational Testing of the General Electric Smart Grid Capable Electric Vehicle Supply Equipment (EVSE)

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Richard Barney [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scoffield, Don [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bennett, Brion [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-12-01

    The Idaho National Laboratory conducted testing and analysis of the General Electric (GE) smart grid capable electric vehicle supply equipment (EVSE), which was a deliverable from GE for the U.S. Department of Energy FOA-554. The Idaho National Laboratory has extensive knowledge and experience in testing advanced conductive and wireless charging systems though INL’s support of the U.S. Department of Energy’s Advanced Vehicle Testing Activity. This document details the findings from the EVSE operational testing conducted at the Idaho National Laboratory on the GE smart grid capable EVSE. The testing conducted on the EVSE included energy efficiency testing, SAE J1772 functionality testing, abnormal conditions testing, and charging of a plug-in vehicle.

  11. Electric arc apparatus for severing split-pin assemblies of guide tubes of nuclear reactors

    International Nuclear Information System (INIS)

    Burns, D.C.; Kauric, C.E.; Persang, J.C.

    1987-01-01

    This patent describes an apparatus for use in the replacement of an old split-pin assembly of a guide tube of a nuclear reactor by a new split-pin assembly, the old split-pin assembly including an old split pin and an old nut securing the old split pin to the guide tube, the old split-pin assembly and the guide tube being radioactive. The apparatus includes a metal disintegration machining tool, the tool having an electrode, means for mounting the tool submerged in a pool of water in engagement with the guide tube and with the old split-pin assembly secured to the guide tube, the tool being so mounted with the electrode in position to coact electrically with the last-named old split-pin assembly but not with the guide tube, and means, connected to the tool, for firing a disintegrating arc between the electrode and the assembly to disintegrate the assembly into readily removable fragments

  12. Steel fibre concrete, a safer material for reactor construction. A general theory for rupture prediction

    International Nuclear Information System (INIS)

    Rammant, J.P.; Van Laethem, L.; Backx, E.

    1977-01-01

    The effect of steel fibre reinforcement on the mechanical behavior of concrete reactor structures is studied. It is shown that this material leads to a higher safety factor for highly stressed concrete structures like prestressed concrete pressure vessels. The reinforcement of concrete with short steel fibres results clearly in a fundamental change of the material properties. The study comprises basic experiments, the elaboration of an expression of the material laws, the development of a general computer program and the comparison of computational results with more elaborate experiments. Basic experimental work is conducted to determine the material characteristics of the fibre reinforced concrete. It is shown how the fibre reinforcement mechanism is translated into mathematical formulae by expressing the principal characteristics as matrix relationships. These relationships describe the elasto-plastic behavior and the cracked behavior. Probabilistic principles are used to express to fibre efficiency, such that a general stress-strain relationship is incorporated in a subsequent computer program. A general finite element program is developed which includes the new matrix relationships, the pull-out of fibres and the general stress-strain equations. A nonlinear calculation method gives the propagation of the distributed cracks with increasing load untill failure of the structure. Similarly, thermal cycling conditions are accounted for. For example the crack propagation in a fibre reinforced beam was measured by the photostress coating technique: the comparison with the computed crack propagation reveals an excellent agreement. Other comparative studies on simple structural parts are also reported

  13. Electric emissions from electrical appliances

    International Nuclear Information System (INIS)

    Leitgeb, N.; Cech, R.; Schroettner, J.

    2008-01-01

    Electric emissions from electric appliances are frequently considered negligible, and standards consider electric appliances to comply without testing. By investigating 122 household devices of 63 different categories, it could be shown that emitted electric field levels do not justify general disregard. Electric reference values can be exceeded up to 11-fold. By numerical dosimetry with homogeneous human models, induced intra-corporal electric current densities were determined and factors calculated to elevate reference levels to accounting for reduced induction efficiency of inhomogeneous fields. These factors were found not high enough to allow generally concluding on compliance with basic restrictions without testing. Electric appliances usually simultaneously emit both electric and magnetic fields exposing almost the same body region. Since the sum of induced current densities is limited, one field component reduces the available margin for the other. Therefore, superposition of electric current densities induced by either field would merit consideration. (authors)

  14. 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

  15. EPR (European Pressurized Reactor)

    International Nuclear Information System (INIS)

    2015-01-01

    This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

  16. 77 FR 65765 - Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General...

    Science.gov (United States)

    2012-10-30

    ... the vehicle. The antenna module translates the radio frequency signal received from the key into a... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General Motors Corporation AGENCY...

  17. 77 FR 25534 - Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General...

    Science.gov (United States)

    2012-04-30

    ... response back to the vehicle. The antenna module translates the radio frequency signal received from the... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General Motors Corporation AGENCY...

  18. 78 FR 77790 - Petition for Exemption From the Federal Motor Vehicle Theft Prevention Standard; General Motors...

    Science.gov (United States)

    2013-12-24

    ... vehicle. The antenna module translates the radio frequency signal received from the key into a digital... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Petition for Exemption From the Federal Motor Vehicle Theft Prevention Standard; General Motors Corporation AGENCY: National...

  19. 9th international conference on high-temperature reactors - coal and nuclear energy for electricity and gas generation

    International Nuclear Information System (INIS)

    Kelber, G.

    1987-01-01

    The site of the high-temperatur reactor in the Ruhr region neighbouring on a coal-fired power plant is not accidental. The potential of the high-temperature reactor as a central plant element for the supply of heat for heating purposes and process heat covers also the possibility of coal gasification and liquefaction. Therefore the high-temperature reactor is, in the long term, a ray of hope for the coal region, able to compensate for the production-related competitive disadvantages of local coal. It can contribute to guaranteeing in the long term the task of German hard coal as an essential pillar of our energy supply. The VGB as a technical association of thermal power plant operators is particularly committed to the integration of coal and nuclear energy. Within the bounds of its possibilities, it will contribute to promoting the safe and environmentally beneficial generation of electricity from the two primary energy sources. (orig./DG) [de

  20. Projected electricity savings from implementing minimum energy efficiency standard for household refrigerators in Malaysia

    International Nuclear Information System (INIS)

    Mahlia, T.M.I.; Masjuki, H.H.; Saidur, R.; Choudhury, I.A.; NoorLeha, A.R.

    2003-01-01

    The Malaysian economy has grown rapidly in the last two decades. This growth has increased the ownership of household electrical appliances, especially refrigerator-freezers. Almost every house in Malaysia owns a refrigerator-freezer. The Malaysia Energy Center considered implementing a minimum energy efficiency standard for household refrigerator-freezers sometime in the coming year. This paper attempts to predict the amount of energy savings in the residential sector by implementing a minimum energy efficiency standard for household refrigerator-freezers. The calculations are based on the growth of refrigerator-freezer ownership data in Malaysian households. By implementing the programs in 2004, about 8722 GWh will be saved in the year 2013. Therefore, efficiency improvement of this appliance will provide a significant impact in future electricity consumption in Malaysia

  1. Hot topic - fire resistant clothing for electrical workers has no Canadian standard

    Energy Technology Data Exchange (ETDEWEB)

    Douglas, R

    1996-07-01

    Fabrics which could be used for protective clothing for electrical workers was discussed. In Canada there are no stringent clothing rules for the electrical industry, but in the United States, the Federal Occupational Safety and Health Administration (OSHA) have made protective clothing an issue within the industry. Dupont`s Nomex, Westex`s Indura, and Hoechst Celanese`s PBI fibres are among the materials which have been through a series of thermal protection tests and are based on proven materials which are now being applied to work wear. The testing standards cover the ignition of the basic materials, untreated cottons, wools, and the thermal protection of flame resistant materials. The final tests are based on high current arc tests conducted on the fabrics used in the industry. These tests are now being conducted at Ontario Hydro`s High Current Test Facility preliminary to a Canadian standard being formulated and proposed in the very near future. 2 figs.

  2. Insulating oil, electrical for transformers and switches : a national standard of Canada

    International Nuclear Information System (INIS)

    Paniri, S.; Burford, G.; Martin, A.; Adragna, M.

    1997-01-01

    Standard specifications for insulating oil used in power transformers, instrument transformers, bushings, bulk oil circuit breakers, oil circuit reclosers, and switches were provided. The specifications are divided into Class A and Class B depending on the requirement for kinematic viscosity at -40 degrees C. A Class S oil is also introduced for oil circuit breakers. The standards were prepared by the Technical Committee on Transformer and Switch Oils under the jurisdiction of the Steering Committee on Electrical Engineering, and has been formally approved by these committees. It has been also approved as a National Standard of Canada by the Standards Council of Canada. The document provides a list of reference publications, describes the samples and test procedures, properties and delivery requirements. 1 tab

  3. Insulating oil, electrical for transformers and switches : a national standard of Canada; 5. ed.

    Energy Technology Data Exchange (ETDEWEB)

    Paniri, S; Burford, G; Martin, A; Adragna, M [eds.

    1997-09-01

    Standard specifications for insulating oil used in power transformers, instrument transformers, bushings, bulk oil circuit breakers, oil circuit reclosers, and switches were provided. The specifications are divided into Class A and Class B depending on the requirement for kinematic viscosity at -40 degrees C. A Class S oil is also introduced for oil circuit breakers. The standards were prepared by the Technical Committee on Transformer and Switch Oils under the jurisdiction of the Steering Committee on Electrical Engineering, and has been formally approved by these committees. It has been also approved as a National Standard of Canada by the Standards Council of Canada. The document provides a list of reference publications, describes the samples and test procedures, properties and delivery requirements. 1 tab.

  4. General method for calculating polarization electric fields produced by auroral Cowling mechanism and application examples

    Science.gov (United States)

    Vanhamäki, Heikki; Amm, Olaf; Fujii, Ryo; Yoshikawa, Aki; Ieda, Aki

    2013-04-01

    The Cowling mechanism is characterized by the generation of polarization space charges in the ionosphere in consequence of a partial or total blockage of FAC flowing between the ionosphere and the magnetosphere. Thus a secondary polarization electric field builds up in the ionosphere, which guarantees that the whole (primary + secondary) ionospheric current system is again in balance with the FAC. In the Earth's ionosphere the Cowling mechanism is long known to operate in the equatorial electrojet, and several studies indicate that it is important also in auroral current systems. We present a general method for calculate the secondary polarization electric field, when the ionospheric conductances, the primary (modeled) or the total (measured) electric field, and the Cowling efficiency are given. Here the Cowling efficiency is defined as the fraction of the divergent Hall current canceled by secondary Pedersen current. In contrast to previous studies, our approach is a general solution which is not limited to specific geometrical setups (like an auroral arc), and all parameters may have any kind of spatial dependence. The solution technique is based on spherical elementary current (vector) systems (SECS). This way, we avoid the need to specify explicit boundary conditions for the searched polarization electric field or its potential, which would be required if the problem was solved in a differential equation approach. Instead, we solve an algebraic matrix equation, for which the implicit boundary condition that the divergence of the polarization electric field vanishes outside our analysis area is sufficient. In order to illustrate the effect of Cowling mechanism on ionospheric current systems, we apply our method to two simple models of auroral electrodynamic situations: 1) a mesoscale strong conductance enhancement in the early morning sector within a relatively weak southward primary electric field, 2) a morning sector auroral arc with only a weak conductance

  5. General layout of a 1300 MW PWR buildings: standard lay-out and influence factors

    International Nuclear Information System (INIS)

    Silva Messias, M. da

    1986-01-01

    The Standardization concept of a 1300 MW Nuclear Power Plant equipped with pressurized-water reactors (PWR) will be considered with regard to the factors which have various degrees of influence on the Layout of the Site. Standardization with regard to the Layout of nuclear power plant, stands, however also for a fixed arrangement of the components in the main buildings and a fixed arrangement of the buildings in the site plan. Standardization may result in simplified and shorter licensing procedures and mean less planning effort. Concentration of systems, optimization of the electro-mechanical design, clear separation of the controlled area, need for shielding, maintenance and safety aspects are some of number of factors which have various degress of influence on the layout of the buildings. (Author) [pt

  6. 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...

  7. Standardization for electric energy saving; Normalizacion para el ahorro de energia electrica

    Energy Technology Data Exchange (ETDEWEB)

    Reyes Diaz, Arturo [Fideicomiso de Apoyo al Programa de Ahorro de Energia del Sector Electrico (FIDE), Mexico, D. F. (Mexico)

    1994-12-31

    This work describes shortly the current legislation on the standardization; as well as the participation of the entities Fideicomiso de Apoyo al Programa de Ahorro de Energia del Sector Electrico (FIDE) and Programa de Ahorro de Energia (PAESE) in the consulting committees, for the elaboration of the Mexican official standards focussed to electric power saving. [Espanol] El presente trabajo, describe brevemente la legislacion actual sobre la normalizacion; asi como la participacion de las entidades Fideicomiso de Apoyo al Programa de Ahorro de Energia del Sector Electrico (FIDE) y Programa de ahorro de Energia (PAESE) en los comites consultivos, para la elaboracion de normas oficiales mexicanas enfocadas al ahorro de energia electrica.

  8. Standardization for electric energy saving; Normalizacion para el ahorro de energia electrica

    Energy Technology Data Exchange (ETDEWEB)

    Reyes Diaz, Arturo [Fideicomiso de Apoyo al Programa de Ahorro de Energia del Sector Electrico (FIDE), Mexico, D. F. (Mexico)

    1993-12-31

    This work describes shortly the current legislation on the standardization; as well as the participation of the entities Fideicomiso de Apoyo al Programa de Ahorro de Energia del Sector Electrico (FIDE) and Programa de Ahorro de Energia (PAESE) in the consulting committees, for the elaboration of the Mexican official standards focussed to electric power saving. [Espanol] El presente trabajo, describe brevemente la legislacion actual sobre la normalizacion; asi como la participacion de las entidades Fideicomiso de Apoyo al Programa de Ahorro de Energia del Sector Electrico (FIDE) y Programa de ahorro de Energia (PAESE) en los comites consultivos, para la elaboracion de normas oficiales mexicanas enfocadas al ahorro de energia electrica.

  9. Magnetohydrodynamic flow of generalized Maxwell fluids in a rectangular micropump under an AC electric field

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Guangpu [School of Mathematical Science, Inner Mongolia University, Hohhot, Inner Mongolia 010021 (China); Jian, Yongjun, E-mail: jianyj@imu.edu.cn [School of Mathematical Science, Inner Mongolia University, Hohhot, Inner Mongolia 010021 (China); Chang, Long [School of Mathematics and Statistics, Inner Mongolia University of Finance and Economics, Hohhot, Inner Mongolia 010051 (China); Buren, Mandula [School of Mathematical Science, Inner Mongolia University, Hohhot, Inner Mongolia 010021 (China)

    2015-08-01

    By using the method of separation of variables, an analytical solution for the magnetohydrodynamic (MHD) flow of the generalized Maxwell fluids under AC electric field through a two-dimensional rectangular micropump is reduced. By the numerical computation, the variations of velocity profiles with the electrical oscillating Reynolds number Re, the Hartmann number Ha, the dimensionless relaxation time De are studied graphically. Further, the comparison with available experimental data and relevant researches is presented. - Highlights: • MHD flow of the generalized Maxwell fluids under AC electric field is analyzed. • The MHD flow is confined to a two-dimensional rectangular micropump. • Analytical solution is obtained by using the method of separation of variables. • The influences of related parameters on the MHD velocity are discussed.

  10. Magnetohydrodynamic flow of generalized Maxwell fluids in a rectangular micropump under an AC electric field

    International Nuclear Information System (INIS)

    Zhao, Guangpu; Jian, Yongjun; Chang, Long; Buren, Mandula

    2015-01-01

    By using the method of separation of variables, an analytical solution for the magnetohydrodynamic (MHD) flow of the generalized Maxwell fluids under AC electric field through a two-dimensional rectangular micropump is reduced. By the numerical computation, the variations of velocity profiles with the electrical oscillating Reynolds number Re, the Hartmann number Ha, the dimensionless relaxation time De are studied graphically. Further, the comparison with available experimental data and relevant researches is presented. - Highlights: • MHD flow of the generalized Maxwell fluids under AC electric field is analyzed. • The MHD flow is confined to a two-dimensional rectangular micropump. • Analytical solution is obtained by using the method of separation of variables. • The influences of related parameters on the MHD velocity are discussed

  11. IEEE 1547 and 2030 Standards for Distributed Energy Resources Interconnection and Interoperability with the Electricity Grid

    Energy Technology Data Exchange (ETDEWEB)

    Basso, T.

    2014-12-01

    Public-private partnerships have been a mainstay of the U.S. Department of Energy and the National Renewable Energy Laboratory (DOE/NREL) approach to research and development. These partnerships also include technology development that enables grid modernization and distributed energy resources (DER) advancement, especially renewable energy systems integration with the grid. Through DOE/NREL and industry support of Institute of Electrical and Electronics Engineers (IEEE) standards development, the IEEE 1547 series of standards has helped shape the way utilities and other businesses have worked together to realize increasing amounts of DER interconnected with the distribution grid. And more recently, the IEEE 2030 series of standards is helping to further realize greater implementation of communications and information technologies that provide interoperability solutions for enhanced integration of DER and loads with the grid. For these standards development partnerships, for approximately $1 of federal funding, industry partnering has contributed $5. In this report, the status update is presented for the American National Standards IEEE 1547 and IEEE 2030 series of standards. A short synopsis of the history of the 1547 standards is first presented, then the current status and future direction of the ongoing standards development activities are discussed.

  12. Perceived benefits of adopting Standard – Based pricing mechanism for mechanical and electrical services installations

    OpenAIRE

    Ganiyu Amuda Yusuf; Sarajul Fikri Mohamed

    2014-01-01

    Cost is an important measure of project success and clients will expect a reliable forecast at the early stage of construction projects to inform their business decision. This study was undertaken to investigate the current practices in managing cost of mechanical and electrical (M&E) services in buildings. The perceptions of practitioners on the benefits of adopting Standard – Based Pricing Mechanism for M&E services as used for building fabrics and finishes was ascertained. The meth...

  13. 76 FR 12792 - Petition for Exemption From the Federal Motor Vehicle Motor Theft Prevention Standard; General...

    Science.gov (United States)

    2011-03-08

    ... From the Federal Motor Vehicle Motor Theft Prevention Standard; General Motors Corporation AGENCY..., Exemption from the Theft Prevention Standard. This petition is granted because the agency has determined... in reducing and deterring motor vehicle theft as compliance with the parts-marking requirements of...

  14. Review of electricity supply failures and plant improvements over 25 years operation of the Harwell materials test reactors

    International Nuclear Information System (INIS)

    Taylor, D.J.

    1986-01-01

    The evolution of the on-site electrical power sources is described, operational experience is reported and shortcomings are identified. Disturbances in the external power supplies to the reactors are listed for the past 25 years and failure probabilities are derived from this historical data. The 132 kV overhead supply to the Harwell site is identified as the source of nearly 90% of the disturbances. (author)

  15. Review of electricity supply failures and plant improvements over 25 years operation of the Harwell materials test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, D. J. [UKAEA Harwell (United Kingdom)

    1986-02-15

    The evolution of the on-site electrical power sources is described, operational experience is reported and shortcomings are identified. Disturbances in the external power supplies to the reactors are listed for the past 25 years and failure probabilities are derived from this historical data. The 132 kV overhead supply to the Harwell site is identified as the source of nearly 90% of the disturbances. (author)

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

    International Nuclear Information System (INIS)

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

    1982-06-01

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

  17. 76 FR 68634 - Airworthiness Directives; General Electric Company (GE) CF6 Turbofan Engines

    Science.gov (United States)

    2011-11-07

    ... Airworthiness Directives; General Electric Company (GE) CF6 Turbofan Engines AGENCY: Federal Aviation... ``(c) This AD applies to * * * and CF6-80A3 turbofan engines with left-hand links * * *.'' to ``(c) This AD applies to * * * and CF6-80A3 turbofan engines, including engines marked on the engine data...

  18. 75 FR 47644 - General Electric Company, Transportation Division, Including On-Site Leased Workers From Adecco...

    Science.gov (United States)

    2010-08-06

    ... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-71,174] General Electric Company... that was totally or partially separated from employment. It was determined, however, that imports of... Globalization Adjustment Assistance Act of 2009, section 222 of the Act (19 U.S.C. 2272) covers foreign...

  19. Hazardous Waste Cleanup: Caribe General Electric Product in Rio Piedras, Puerto Rico

    Science.gov (United States)

    The Caribe General Electric Product facility (GE Río Piedras) covers approximately 4 acres and is situated on the northern coastal plain in a relatively flat industrial/urban area of Río Piedras with an elevation between 20 and 40 feet above mean sea leve

  20. The Impact of Minimum Energy Performance Standards (MEPS) Regulation on Electricity Saving in Malaysia

    Science.gov (United States)

    Fatihah Salleh, Siti; Eqwan Roslan, Mohd; Isa, Aishah Mohd; Faizal Basri Nair, Mohd; Syafiqah Salleh, Siti

    2018-03-01

    One of Malaysia’s key strategies to promote efficient energy use in the country is to implement the minimum energy performance standards (MEPS) through the Electricity Regulations (Amendment) 2013. Five selected electrical appliances (refrigerator, air conditioner, television, domestic fans and lamp fittings) must comply with MEPS requirement in order to be sold in Malaysian market. Manufacturers, importers or distributors are issued Certificate of Approval (COA) if products are MEPS-compliant. In 2015, 1,215 COAs were issued but the number of MEPS products in the market is unknown. This work collects sales data from major manufacturers to estimate the annual sales of MEPS appliances and the cumulative electricity consumption and electricity saving. It was found that most products sold have 3-star rating and above. By year 2015, total cumulative electricity savings gained from MEPS implementation is 3,645 GWh, with air conditioner being the highest contributor (30%). In the future, it is recommended that more MEPS products and related incentives be introduced to further improve efficiency of energy use in Malaysia.

  1. US Advanced Light Water Reactor Program; overall objective

    International Nuclear Information System (INIS)

    Klug, N.

    1989-01-01

    The overall objective of the US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) program is to perform coordinated programs of the nuclear industry and DOE to insure the availability of licensed, improved, and simplified light water reactor standard plant designs that may be ordered in the 1990's to help meet the US electrical power demand. The discussion includes plans to meet program objectives and the design certification program. DOE is currently supporting the development of conceptual designs, configurations, arrangements, construction methods/plans, and proof test key design features for the General Electric ASBWR and the Westinghouse AP600. Key features of each are summarized. Principal milestones related to licensing of large standard plants, simplified mid-size plant development, and plant lifetime improvement are noted

  2. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko

    2000-03-01

    A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

  3. 42 CFR 403.206 - General standards for Medicare supplemental policies.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 2 2010-10-01 2010-10-01 false General standards for Medicare supplemental policies. 403.206 Section 403.206 Public Health CENTERS FOR MEDICARE & MEDICAID SERVICES, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL PROVISIONS SPECIAL PROGRAMS AND PROJECTS Medicare Supplemental Policies...

  4. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. General synthesis

    International Nuclear Information System (INIS)

    Hery, M.; Lecocq, A.

    1983-03-01

    After a brief recall of the MSBR project, French studies on molten salt reactors are summed up. Theoretical and experimental studies for a graphite moderated 1000 MWe reactor using molten Li, Be, Th and U fluorides cooled by salt-lead direct contact are given. These studies concern the core, molten salt chemistry, graphite, metals (molybdenum, alloy TZM), corrosion, reactor components [fr

  5. Procedure for qualification of electric equipment installed in containments for pressurized water reactors subject to accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    This generic norm is usable for electrical equipment installed in containment building of PWR subject to accidental conditions. She defines the qualification methods and the general rules usable for the test specifications of qualification for these materials

  6. Examination on the testing method for evaluating life of electric wires and cables for nuclear reactors

    International Nuclear Information System (INIS)

    Seguchi, Tadao; Morita, Yosuke; Yoshida, Kenzo

    1984-01-01

    Regarding the method of environmental test on the electric wires and cables used for the safety system in the containment vessels of nuclear power plants, the draft recommendation was issued by the Institute of Electrical Engineers of Japan in 1982. Its contents follow the IEEE Standard of USA, and are composed of the tests on the deterioration in normal operation, the case of LOCA and the prevention of spread of fire. In this report, as to the testing method regarding normal operation, the appropriate method and its basis are described in view of the recent experimental data. In the draft recommendation, the successive method carrying out irradiation after thermal deterioration is adopted, and both testing conditions are given. However, the fundamental problems remain in the propriety of the acceleration of deterioration and the multiplied effect of heat and radiation. The qualitative and quantitative data on these problems have been accumulated in various countries, therefore, the examination of the testing method was carried out based on these data. The dose rate dependence of radiation deterioration, the multiplied effect of radiation and heat, and the correlation of thermal deterioration rate with temperature are discussed. The appropriate method is proposed. (Kako, I.)

  7. Installation modification of the reactor No.2 of Ikata nuclear power plant of Shikoku Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1980-01-01

    The application was made on August 25, 1979, from the president of the Shikoku Electric Power Company, Inc., to the Minister of International Trade and Industry, relating to the installation modification of the reactor No. 2 in the Ikata nuclear power plant. The inquiry was submitted on September 28, 1979, from the Minister of International Trade and Industry to the Nuclear Safety Commission, after the safety evaluation in the Ministry of International Trade and Industry, and the investigation and deliberation were started on October 1, 1979, in the Nuclear Safety Commission. The content of the modification is to add the circuit actuated by the abnormal low pressure signal of the reactor to the actuating circuit of the emergency core cooling system (ECCS) and to increase the new fuel storage capacity from about 1/3 core to about 2/3 core. The additional signal circuit is composed of the logic circuit of ''2 out of 4'' and is multichannel design. The circuit is independent from the reactor control system and the conventional signal circuit of the concurrence of low pressure in the reactor and low level in the pressurizer. With the addition of the circuit of abnormal low pressure signal of the reactor, the countermeasures for preventing ECCS start by mistake are also added. These modifications give no influence to the functions of the reactor control system and reactor protection system. The function and the performance of ECCS were analyzed and evaluated accompanying these modifications assuming the loss of coolant accident. Concerning the new fuel storage capacity, the type of racks is modified from angle type to can type, and the subcriticality is kept even at the time of water flood. (Nakai, Y.)

  8. 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

  9. Standard Test Method for Electrical Performance of Photovoltaic Cells Using Reference Cells Under Simulated Sunlight

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers the determination of the electrical performance of a photovoltaic cell under simulated sunlight by means of a calibrated reference cell procedure. 1.2 Electrical performance measurements are reported with respect to a select set of standard reporting conditions (SRC) (see Table 1) or to user-specified conditions. 1.2.1 The SRC or user-specified conditions include the cell temperature, the total irradiance, and the reference spectral irradiance distribution. 1.3 This test method is applicable only to photovoltaic cells with a linear response over the range of interest. 1.4 The cell parameters determined by this test method apply only at the time of test, and imply no past or future performance level. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 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 s...

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

    International Nuclear Information System (INIS)

    1989-02-01

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

  11. South Africa's nuclear model: A small and innovative reactor is seen as the model for new electricity plants. The project is nearing the starting blocks

    International Nuclear Information System (INIS)

    Ferreira, Tom

    2004-01-01

    Although nuclear power generation has by far the best safety and environmental record of any technology in general use, it has for many years been unable to make any meaningful inroads into the wall of negative perceptions that have arisen against it. But sentiments are changing rapidly on a global scale. The flare-up of oil prices is a sobering reminder of the volatility in the energy market, the exhaustibility of fossil fuels and the urgent need for stable, reliable, non-polluting sources of electrical power that are indispensable to a modern industrial economy. Today, new types of nuclear plants are prized, and South Africa is moving ahead. The State energy provider, Eskom, is internationally regarded as the leader in the field of the Pebble Bed Modular Reactor (PBMR) technology, a 'new generation' nuclear power plant. A decision on the PBMR project's future is on the near horizon. Should approvals be received in the coming months to proceed to the project's next phase, construction of the PBMR demonstration plant will start in 2006, in which case the reactor will start in 2010 and handed over to the client, Eskom, in 2011. Eskom has conditionally undertaken to purchase the first commercial units. Pebble bed reactors are small, about one-sixth the size of most current nuclear plants. Multiple PBMRs can share a common control center and occupy an area of no more than three football fields. More specifically, the PBMR is a helium-cooled, graphite moderated high temperature reactor (HTR). The concept is based on experience in the UK, United States and particularly Germany where prototype reactors were operated successfully between the late 1960s and 1980s. Although it is not the only high-temperature, gas-cooled nuclear reactor being developed in the world, the South African project is internationally regarded as a front-runner. The South African PBMR includes unique and patented technological innovations which make it particularly competitive. The Chief Executive

  12. Higher order generalized perturbation theory for boiling water reactor in-core fuel management optimization

    International Nuclear Information System (INIS)

    Moore, B.R.; Turinsky, P.J.

    1998-01-01

    Boiling water reactor (BWR) loading pattern assessment requires solving the two-group, nodal form of the neutron diffusion equation and drift-flux form of the fluid equations simultaneously because these equation sets are strongly coupled via nonlinear feedback. To reduce the computational burden associated with the calculation of the core attributes (that is, core eigenvalue and thermal margins) of a perturbed BWR loading pattern, the analytical and numerical aspects of a higher order generalized perturbation theory (GPT) method, which correctly addresses the strong nonlinear feedbacks of two-phase flow, have been established. Inclusion of Jacobian information in the definition of the generalized flux adjoints provides for a rapidly convergent iterative method for solution of the power distribution and eigenvalue of a loading pattern perturbed from a reference state. Results show that the computational speedup of GPT compared with conventional forward solution methods demanding consistent accuracy is highly dependent on the number of spatial nodes utilized by the core simulator, varying from superior to inferior performance as the number of nodes increases

  13. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

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

  14. Supply amount and marginal price of renewable electricity under the renewables portfolio standard in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nishio, Kenichiro; Asano, Hiroshi [Central Research institute of Electric Power Industry, Tokyo (Japan). Socio-economic Research Center

    2006-10-15

    The Renewables Portfolio Standard (RPS) in Japan requires that approximately 1.35% of each retail supplier's electricity sales in FY2010 come from renewable energy sources (RES), for example, photovoltaics, wind, biomass, geothermal, and small hydropower. To help retail suppliers and renewable generators develop effective strategies, this study provides a quantitative analysis of the impact of this measure. We assume the supply conditions for electricity generation from renewable energy sources (RES-E) based on regional resource endowments, and we derive the cost-effective compositions of renewable portfolios, RES-E certificate prices, and additional costs to retail suppliers. The future prospects of RES-E are assessed based on technology, region, and year up to FY2010. The analysis reveals that wind power and biomass power generated from municipal waste will provide the majority of the total supply of RES-E under the RPS. It also indicates that the marginal price of RES-E certificates will be approximately 5.8 JPY/kWh (5.2 USc/kWh) in FY2010, in the case wherein the marginal price of electricity is assumed to be 4 JPY/kWh (3.6 USc/kWh). In order to elaborate on this further, sensitivity analyses for some parameters of RES and the price of electricity are provided. The dynamic supply curves of RES-E certificates are also indicated. (author)

  15. Supply amount and marginal price of renewable electricity under the renewables portfolio standard in Japan

    International Nuclear Information System (INIS)

    Nishio, Kenichiro; Asano, Hiroshi

    2006-01-01

    The Renewables Portfolio Standard (RPS) in Japan requires that approximately 1.35% of each retail supplier's electricity sales in FY2010 come from renewable energy sources (RES), for example, photovoltaics, wind, biomass, geothermal, and small hydropower. To help retail suppliers and renewable generators develop effective strategies, this study provides a quantitative analysis of the impact of this measure. We assume the supply conditions for electricity generation from renewable energy sources (RES-E) based on regional resource endowments, and we derive the cost-effective compositions of renewable portfolios, RES-E certificate prices, and additional costs to retail suppliers. The future prospects of RES-E are assessed based on technology, region, and year up to FY2010. The analysis reveals that wind power and biomass power generated from municipal waste will provide the majority of the total supply of RES-E under the RPS. It also indicates that the marginal price of RES-E certificates will be approximately 5.8 JPY/kWh (5.2 USc/kWh) in FY2010, in the case wherein the marginal price of electricity is assumed to be 4 JPY/kWh (3.6 USc/kWh). In order to elaborate on this further, sensitivity analyses for some parameters of RES and the price of electricity are provided. The dynamic supply curves of RES-E certificates are also indicated

  16. Supply amount and marginal price of renewable electricity under the renewables portfolio standard in Japan

    International Nuclear Information System (INIS)

    Nishio, Kenichiro; Asano, Hiroshi

    2006-01-01

    The Renewables Portfolio Standard (RPS) in Japan requires that approximately 1.35% of each retail supplier's electricity sales in FY2010 come from renewable energy sources (RES), for example, photovoltaics, wind, biomass, geothermal, and small hydropower. To help retail suppliers and renewable generators develop effective strategies, this study provides a quantitative analysis of the impact of this measure. We assume the supply conditions for electricity generation from renewable energy sources (RES-E) based on regional resource endowments, and we derive the cost-effective compositions of renewable portfolios, RES-E certificate prices, and additional costs to retail suppliers. The future prospects of RES-E are assessed based on technology, region, and year up to FY2010. The analysis reveals that wind power and biomass power generated from municipal waste will provide the majority of the total supply of RES-E under the RPS. It also indicates that the marginal price of RES-E certificates will be approximately 5.8 JPY/kWh (5.2 USc/kWh) in FY2010, in the case wherein the marginal price of electricity is assumed to be 4 JPY/kWh (3.6 USc/kWh). In order to elaborate on this further, sensitivity analyses for some parameters of RES and the price of electricity are provided. The dynamic supply curves of RES-E certificates are also indicated. (author)

  17. Electrical Stimulation of the Ventral Tegmental Area Induces Reanimation from General Anesthesia

    Science.gov (United States)

    Solt, Ken; Van Dort, Christa J.; Chemali, Jessica J.; Taylor, Norman E.; Kenny, Jonathan D.; Brown, Emery N.

    2014-01-01

    BACKGROUND Methylphenidate or a D1 dopamine receptor agonist induce reanimation (active emergence) from general anesthesia. We tested whether electrical stimulation of dopaminergic nuclei also induces reanimation from general anesthesia. METHODS In adult rats, a bipolar insulated stainless steel electrode was placed in the ventral tegmental area (VTA, n = 5) or substantia nigra (SN, n = 5). After a minimum 7-day recovery period, the isoflurane dose sufficient to maintain loss of righting was established. Electrical stimulation was initiated and increased in intensity every 3 min to a maximum of 120μA. If stimulation restored the righting reflex, an additional experiment was performed at least 3 days later during continuous propofol anesthesia. Histological analysis was conducted to identify the location of the electrode tip. In separate experiments, stimulation was performed in the prone position during general anesthesia with isoflurane or propofol, and the electroencephalogram was recorded. RESULTS To maintain loss of righting, the dose of isoflurane was 0.9% ± 0.1 vol%, and the target plasma dose of propofol was 4.4 μg/ml ± 1.1 μg/ml (mean ± SD). In all rats with VTA electrodes, electrical stimulation induced a graded arousal response including righting that increased with current intensity. VTA stimulation induced a shift in electroencephalogram peak power from δ (anesthesia with isoflurane or propofol. These results are consistent with the hypothesis that dopamine release by VTA, but not SN, neurons induces reanimation from general anesthesia. PMID:24398816

  18. A general modeling method for I-V characteristics of geometrically and electrically configured photovoltaic arrays

    International Nuclear Information System (INIS)

    Liu Guangyu; Nguang, Sing Kiong; Partridge, Ashton

    2011-01-01

    Highlights: → A novel and general method is proposed for modeling PV arrays or modules. → A robust algorithm is used for the first time to improve the convergence to solution. → Auxiliary functions in other general methods are not compulsory in our method. → It is novel that geometric configuration is also incorporated. → A case study is performed to show the approach's advantages and unique features. - Abstract: A general method for modeling typical photovoltaic (PV) arrays and modules is proposed to find the exact current and voltage relationship of PV arrays or modules of geometrically and electrically different configurations. Nonlinear characteristic equations of electrical devices in solar array or module systems are numerically constructed without adding any virtual electrical components. Then, a robust damped Newton method is used to find exact I-V relationship of these general nonlinear equations, where the convergence is guaranteed. The model can deal with different mismatch effects such as different configurations of bypass diodes, and partial shading. Geometry coordinates of PV components are also considered to facilitate the modeling of the actual physical configuration. Simulation of a PV array with 48 modules, partially shaded by a concrete structure, is performed to verify the effectiveness and advantages of the proposed method.

  19. Asymmetric kinetic equilibria: Generalization of the BAS model for rotating magnetic profile and non-zero electric field

    Science.gov (United States)

    Dorville, Nicolas; Belmont, Gérard; Aunai, Nicolas; Dargent, Jérémy; Rezeau, Laurence

    2015-09-01

    Finding kinetic equilibria for non-collisional/collisionless tangential current layers is a key issue as well for their theoretical modeling as for our understanding of the processes that disturb them, such as tearing or Kelvin Helmholtz instabilities. The famous Harris equilibrium [E. Harris, Il Nuovo Cimento Ser. 10 23, 115-121 (1962)] assumes drifting Maxwellian distributions for ions and electrons, with constant temperatures and flow velocities; these assumptions lead to symmetric layers surrounded by vacuum. This strongly particular kind of layer is not suited for the general case: asymmetric boundaries between two media with different plasmas and different magnetic fields. The standard method for constructing more general kinetic equilibria consists in using Jeans theorem, which says that any function depending only on the Hamiltonian constants of motion is a solution to the steady Vlasov equation [P. J. Channell, Phys. Fluids (1958-1988) 19, 1541 (1976); M. Roth et al., Space Sci. Rev. 76, 251-317 (1996); and F. Mottez, Phys. Plasmas 10, 1541-1545 (2003)]. The inverse implication is however not true: when using the motion invariants as variables instead of the velocity components, the general stationary particle distributions keep on depending explicitly of the position, in addition to the implicit dependence introduced by these invariants. The standard approach therefore strongly restricts the class of solutions to the problem and probably does not select the most physically reasonable. The BAS (Belmont-Aunai-Smets) model [G. Belmont et al., Phys. Plasmas 19, 022108 (2012)] used for the first time the concept of particle accessibility to find new solutions: considering the case of a coplanar-antiparallel magnetic field configuration without electric field, asymmetric solutions could be found while the standard method can only lead to symmetric ones. These solutions were validated in a hybrid simulation [N. Aunai et al., Phys. Plasmas (1994-present) 20

  20. Asymmetric kinetic equilibria: Generalization of the BAS model for rotating magnetic profile and non-zero electric field

    International Nuclear Information System (INIS)

    Dorville, Nicolas; Belmont, Gérard; Aunai, Nicolas; Dargent, Jérémy; Rezeau, Laurence

    2015-01-01

    Finding kinetic equilibria for non-collisional/collisionless tangential current layers is a key issue as well for their theoretical modeling as for our understanding of the processes that disturb them, such as tearing or Kelvin Helmholtz instabilities. The famous Harris equilibrium [E. Harris, Il Nuovo Cimento Ser. 10 23, 115–121 (1962)] assumes drifting Maxwellian distributions for ions and electrons, with constant temperatures and flow velocities; these assumptions lead to symmetric layers surrounded by vacuum. This strongly particular kind of layer is not suited for the general case: asymmetric boundaries between two media with different plasmas and different magnetic fields. The standard method for constructing more general kinetic equilibria consists in using Jeans theorem, which says that any function depending only on the Hamiltonian constants of motion is a solution to the steady Vlasov equation [P. J. Channell, Phys. Fluids (1958–1988) 19, 1541 (1976); M. Roth et al., Space Sci. Rev. 76, 251–317 (1996); and F. Mottez, Phys. Plasmas 10, 1541–1545 (2003)]. The inverse implication is however not true: when using the motion invariants as variables instead of the velocity components, the general stationary particle distributions keep on depending explicitly of the position, in addition to the implicit dependence introduced by these invariants. The standard approach therefore strongly restricts the class of solutions to the problem and probably does not select the most physically reasonable. The BAS (Belmont-Aunai-Smets) model [G. Belmont et al., Phys. Plasmas 19, 022108 (2012)] used for the first time the concept of particle accessibility to find new solutions: considering the case of a coplanar-antiparallel magnetic field configuration without electric field, asymmetric solutions could be found while the standard method can only lead to symmetric ones. These solutions were validated in a hybrid simulation [N. Aunai et al., Phys. Plasmas (1994-present