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Sample records for carolina power light robinson-2 reactor

  1. 75 FR 3942 - Carolina Power & Light Company Shearon Harris Nuclear Power Plant, Unit 1 Environmental...

    Science.gov (United States)

    2010-01-25

    ... impact (Part 73, Power Reactor Security Requirements, 74 FR 13926 through 13967, dated March 27, 2009... Carolina Power & Light Company (the licensee), now doing business as Progress Energy Carolinas, Inc. (PEC... promulgating its revisions to 10 CFR Part 73 as discussed in a Federal Register (FR) notice dated March...

  2. 75 FR 80547 - Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit No. 1; Exemption

    Science.gov (United States)

    2010-12-22

    ... COMMISSION Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit No. 1; Exemption 1.0... License No. NPF-63, which authorizes operation of the Shearon Harris Nuclear Power Plant (HNP), Unit 1... nuclear power reactors against radiological sabotage,'' published as a final rule in the Federal...

  3. 75 FR 9958 - Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit 1; Exemption

    Science.gov (United States)

    2010-03-04

    ... COMMISSION Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit 1; Exemption 1.0... of the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The license provides, among other things... for physical protection of licensed activities in nuclear power reactors against radiological...

  4. Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.

    1984-08-01

    A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.

  5. Refined Monte Carlo analysis of the H.B. Robinson-2 reactor pressure vessel dosimetry benchmark

    International Nuclear Information System (INIS)

    Highlights: → Activation of in- and ex-vessel radiometric dosimeters is studied with MCNPX. → Influences of neutron source definition and cross-section libraries are examined. → 237Np(n,f) energy cut-off is set at 10 eV to cover the reaction completely. → Different methods for deriving activities from reaction rates are compared. → Uncertainties are evaluated and are below 10%, final C/E ratios being within 15%. - Abstract: Refined analysis, based on use of the Monte Carlo code MCNPX-2.4.0, is presented for the 'H.B. Robinson-2 pressure vessel dosimetry benchmark', which is a part of the Radiation Shielding and Dosimetry Experiments Database (SINBAD). First, the performance of the Monte Carlo methodology is reassessed relative to the reported deterministic results obtained with DORT. Thereby, the analysis is accompanied by a quantitative evaluation of the optimal energy cut-off value for each of the in- and ex-vessel dosimeters that were employed. Second, a more realistic definition of the neutron source is implemented than proposed in the benchmark. Thus, the current procedure for power-to-neutron-source-strength conversion, as also for explicitly considering the burnup-dependent fuel assembly-wise average fission neutron spectrum, is found to affect the calculated values significantly. In addition to the modelling refinements made, different approaches are tested for deriving the dosimeter activities, such that the neutron source time-evolution and the activity decay can be taken into account more accurately. Finally, in order to achieve a certain assessment of uncertainties, several sensitivity studies are carried out, e.g. with respect to the nuclear data used for the dosimeters, as also to the assumed physical location of the dosimeters. In spite of some apparent degradation in the prediction of experimental results when refining the Monte Carlo modelling, the final calculation/experiment (C/E) ratios for the measured dosimeter activities remain

  6. 75 FR 16871 - Carolina Power & Light Company, Brunswick Steam Electric Plant, Units 1 and 2; Exemption

    Science.gov (United States)

    2010-04-02

    ... COMMISSION Carolina Power & Light Company, Brunswick Steam Electric Plant, Units 1 and 2; Exemption 1.0 Background Carolina Power & Light Company (CP&L, the licensee) is the holder of Facility Operating Renewed License Nos. DPR-71 and DPR-62, which authorize operation of the Brunswick Steam Electric Plant...

  7. 77 FR 8903 - Environmental Assessment and Finding of No Significant Impact; Carolina Power and Light Company...

    Science.gov (United States)

    2012-02-15

    ..., Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory...\\ alloy fuel rod cladding for Renewed Facility Operating License No. NPF-63, issued to Carolina Power and... 50.46, paragraph (a)(1)(i) provides requirements for reactors containing uranium oxide fuel...

  8. 77 FR 13156 - Carolina Power & Light Company; Shearon Harris Nuclear Power Plant, Unit 1; Exemption

    Science.gov (United States)

    2012-03-05

    ... reactors,'' requires, among other items, that each boiling or pressurized light-water nuclear power reactor... 50, Appendix K, make no provisions for use of fuel rods clad in a material other than zircaloy or... Appendix K to 10 CFR part 50 in order to use the fuel rods clad with AREVA's M5 \\TM\\ alloy. 3.0...

  9. 76 FR 68512 - Carolina Power & Light Company; H. B. Robinson Steam Electric Plant, Unit 2; Exemption

    Science.gov (United States)

    2011-11-04

    ... applicable to reactors fueled with M5 fuel rod cladding. The NRC staff's review and approval of topical... reactors fueled with zircaloy clad fuel are also applicable to reactors fueled with M5 fuel rod cladding... cooling systems for light- water nuclear power reactors,'' paragraph (a)(1)(i) provides requirements...

  10. 75 FR 82414 - Carolina Power & Light Company; H. B. Robinson Steam Electric Plant, Unit No. 2; Exemption

    Science.gov (United States)

    2010-12-30

    ... COMMISSION Carolina Power & Light Company; H. B. Robinson Steam Electric Plant, Unit No. 2; Exemption 1.0 Background Carolina Power & Light Company (CP&L, the licensee) is the holder of Renewed Facility Operating License No. DPR-23, which authorizes operation of the H.B. Robinson Steam Electric Plant, Unit 2...

  11. 75 FR 11579 - Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2; Exemption

    Science.gov (United States)

    2010-03-11

    ... COMMISSION Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2; Exemption 1.0 Background Carolina Power & Light Company (the licensee) is the holder of Facility Operating License No. DPF-23, which authorizes operation of the H. B. Robinson Steam Electric Plant, Unit 2 (HBRSEP)....

  12. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  13. 76 FR 66333 - Carolina Power & Light Company, H.B. Robinson Steam Electric Plant, Unit No. 2; Environmental...

    Science.gov (United States)

    2011-10-26

    ..., ``Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,'' and 10 CFR part 50, appendix k, ``ECCS Evaluation Models,'' to allow for the use of M5 alloy fuel rod cladding for... regulations either state that either zircaloy or ZIRLO is used as the fuel rod cladding material. The...

  14. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  15. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... analysis for liquid and gaseous radwaste system components for light water nuclear power reactors... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections...

  16. Preliminary development of an integrated approach to the evaluation of pressurized thermal shock as applied to the Oconee Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    An evaluation of the risk to the Oconee-1 nuclear plant due to pressurized thermal shock (PTS) has been Completed by Oak Ridge National Laboratory (ORNL). This evaluaion was part of a Nuclear Regulatory Commission (NRC) program designed to study the PTS risk to three nuclear plants: Oconee-1, a Babcock and Wilco reactor plant owned and operated by Duke Power Company; Calvert Cliffs-1, a Combustion Engineering reactor plant owned and operated by Baltimore Gas and Electric company; and H.B. Robinson-2, a Westinghouse reactor plant owned and operated by Carolina Power and Light Company. Studies of Calvert Cliffs-1 and H.B. Robinson-2 are still underway. The specific objectives of the Oconee-1 study were to: (1) provide a best estimate of the probability of a through-the-wall crack (TWC) occurring in the reactor pressure vessel as a result of PTS; (2) determine dominant accident sequences, plant features, operator and control actions and uncertainty in the PTS risk; and (3) evaluate effectiveness of potential corrective measures

  17. 75 FR 8753 - Carolina Power & Light Company, Brunswick Steam Electric Plant, Units 1 and 2; Environmental...

    Science.gov (United States)

    2010-02-25

    ... part 73 as discussed in a Federal Register notice dated March 27, 2009 (74 FR 13967). There will be no... Requirements, 74 FR 13926, 13967 (March 27, 2009)). The licensee currently maintains a security system... Electric Plant, Units 1 and 2 (BSEP), located in Brunswick County, North Carolina. In accordance with...

  18. 77 FR 40092 - License Amendment To Increase the Maximum Reactor Power Level, Florida Power & Light Company, St...

    Science.gov (United States)

    2012-07-06

    ... and limit high temperatures to the mixing zone area specified in the IWFP. The NRC also analyzed the... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION License Amendment To Increase the Maximum Reactor Power Level, Florida Power & Light Company,...

  19. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

  1. Inquiry into the radiological consequences of power uprates at light-water reactors worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Bilic Zabric, Tea; Tomic, Bojan; Lundgren, Klas; Sjoeberg, Mats

    2007-05-15

    In Sweden, most of the nuclear power plants are planning power uprates within the next few years. The Dept. of Occupational and Medical Exposures at the Swedish Radiation Protection Agency, SSI, has initiated a research project to investigate the radiological implications of power uprates on light-water reactors throughout the world. The project was divided into three tasks: 1. A compilation of power uprates of light-water reactors worldwide. The compilation contains a technical description in brief of how the power uprates were carried out. 2. An analysis of the radiological consequences at four selected Nuclear Power Plants, which was the main objective of the inquiry. Affects on the radiological and chemical situation due to the changed situation were discussed. 3. Review of technical and organisational factors to be considered in uprate projects to keep exposures ALARA. The project was carried out, starting with the collecting of information on the implemented and planned uprates on reactors internationally. The information was catalogued in accordance with criteria focusing on radiological impact. A detailed analysis followed of four plants selected for uprates chosen according to established criteria, in line with the project requirements. The selected plants were Olkiluoto 1 and 2, Cofrentes, Asco and Tihange. The plants were selected with design and operation conditions close to the Swedish plants. All information was compiled to identify good and bad practices that are impacting on the occupational exposure. Important factors were discussed concerning BWRs and PWRs which affect radiation levels and occupational exposures in general, and especially at power uprates. Conclusions related to each task are in detail presented in a particular chapter of the report. Taking into account the whole project and its main objective the following conclusions are considered to be emphasized: Optimisation of the work processes to limit the duration of the time spent in

  2. Inquiry into the radiological consequences of power uprates at light-water reactors worldwide

    International Nuclear Information System (INIS)

    In Sweden, most of the nuclear power plants are planning power uprates within the next few years. The Dept. of Occupational and Medical Exposures at the Swedish Radiation Protection Agency, SSI, has initiated a research project to investigate the radiological implications of power uprates on light-water reactors throughout the world. The project was divided into three tasks: 1. A compilation of power uprates of light-water reactors worldwide. The compilation contains a technical description in brief of how the power uprates were carried out. 2. An analysis of the radiological consequences at four selected Nuclear Power Plants, which was the main objective of the inquiry. Affects on the radiological and chemical situation due to the changed situation were discussed. 3. Review of technical and organisational factors to be considered in uprate projects to keep exposures ALARA. The project was carried out, starting with the collecting of information on the implemented and planned uprates on reactors internationally. The information was catalogued in accordance with criteria focusing on radiological impact. A detailed analysis followed of four plants selected for uprates chosen according to established criteria, in line with the project requirements. The selected plants were Olkiluoto 1 and 2, Cofrentes, Asco and Tihange. The plants were selected with design and operation conditions close to the Swedish plants. All information was compiled to identify good and bad practices that are impacting on the occupational exposure. Important factors were discussed concerning BWRs and PWRs which affect radiation levels and occupational exposures in general, and especially at power uprates. Conclusions related to each task are in detail presented in a particular chapter of the report. Taking into account the whole project and its main objective the following conclusions are considered to be emphasized: Optimisation of the work processes to limit the duration of the time spent in

  3. Natural uranium fueled light water moderated breeding hybrid power reactors: a feasibility study

    International Nuclear Information System (INIS)

    The first part of the study consists of a thorough investigation of the properties of subcritical thermal lattices for hybrid reactor applications. Light water is found to be the best moderator for (fuel-self-sufficient) FSS hybrid reactors for power generation. Several lattice geometries and compositions of particular promise for LWHRs are identified. Using one of these lattices, fueled with natural uranium, the performance of several concepts of LWHR blankets is investigated, and optimal blanket designs are identified. The effect of blanket coverage efficiency and the feasibility of separating the functions of tritium breeding and of power generation to different blankets are investigated. Optimal iron-water shields for LWHRs are also determined. The performance of generic types of LWHRs is evaluated. The evolution of the blanket properties with burnup is evaluated and fuel management schemes are briefly examined. The feasibility of using the lithium system of the blanket to control the blanket power amplitude and shape is also investigated. A parametric study of the energy balance of LWHR power plants is carried out, and performance parameters expected from LWHRs are estimated. Discussions are given of special features of LWHRs and their fuel cycle

  4. Assessment of the use of extended burnup fuel in light water power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

    1988-02-01

    This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

  5. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  6. Safety, technical and economic objectives of the electric power research institute's advanced light water reactor programme

    International Nuclear Information System (INIS)

    In the early 1980s, the Electric Power Research Institute entered into a programme to develop the safety, technical, operational and economic objectives of the next generation of LWRs. The programme benefited from the support and participation of US and international utilities, the United States Department of Energy, US reactor vendors and architect-engineering firms. This programme provided a foundation for the Nuclear Power Oversight Committee's comprehensive initiative for revitalizing nuclear power in the USA, as set forth in its 'Strategic Plan for Building New Nuclear Power Plants', published in November 1990 and updated annually each November. This Strategic Plan contains 14 'building blocks', each of which is considered essential to constructing a new nuclear plants. Building from the perspective that the Advanced Light Water Reactor (ALWR) should reflect lessons learned from the design, construction and operation of existing nuclear power plants, utilities were surveyed to help identify the cornerstone principles of what was subsequently called the 'ALWR Utility Requirements Document' (URD). The paper provides a review of the objectives underlying these principles, and the design targets that derive from this. Some examples are provided of how these targets are being met in the design of the four ALWR plants currently undergoing United States Nuclear Regulatory Commission licensing certification under United States Code of Federal Regulations 10CFR52. Some areas requiring further investigation are also described. The paper concludes by discussing key challenges to the URD targets and process, in particular with respect to regulatory and economic considerations, and provides brief highlights of steps being taken to respond to these challenges. (author). 2 tabs

  7. Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon [Korea Atomic Energy Research Institute, Irradiated Materials Examination Facility, Taejon (Korea)

    1999-09-01

    The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

  8. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the 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 about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  9. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

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

  10. Light water cooled, high temperature and high performance nuclear power plants concept of once-through coolant cycle, supercritical-pressure, light water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Supercritical-pressure, light water cooled nuclear reactors corresponding to nuclear reactors of once-through boilers, are of theoretical development from LWR. Under supercritical pressure, a steam turbine can be driven directly with cooled water with high enthalpy, as not seen boiling and required for recycling. The reactor has no steam-water separation and recycling systems on comparison with the boiling water type LWR, and is the same once-through type as supercritical-pressure thermal power generation plants. Then, all of cooling water at reactor core are sent to turbine. The reactor has no steam generator, and pressurizer, on comparison with PWR. As it requires no steam-water separator, steam drier, and recycling system on comparison with BWR, it becomes of smaller size and has shape and size nearly equal to those of PWR. And, its control bars can be inserted from upper direction like PWR, and can use its driving system. Here was introduced some concepts on high-temperature and high-performance light water reactor, nuclear power generation using a technology on supercritical-pressure thermal power generation. (G.K.)

  11. Results of R+D of nuclear power plants with WWER-1000 type light-water reactors. Part 2

    International Nuclear Information System (INIS)

    Volume II of the proceedings contains full texts of 32 contributions, all of which fall within the INIS subject scope. The conference was organized to facilitate exchange of knowledge and experience between Czechoslovak organizations engaged in the State Research Project ''Nuclear Power Facilities with 1000 MW Light-Water Reactors'' and domestic suppliers of the requisite equipment. (Z.M.). 45 figs., 9 tabs., 201 refs

  12. Results of R+D of nuclear power plants with WWER-1000 type light-water reactors. Part 1

    International Nuclear Information System (INIS)

    Volume I of the proceedings contains full texts of 44 invited lectures and contributions, all of which fall within the INIS subject scope. The conference was organized to facilitate exchange of knowledge and experience between Czechoslovak organizations engaged in the State Research Project ''Nuclear Power Facilities with 1000 MW Light-Water Reactors'' and domestic suppliers of the requisite equipment. (Z.M.). 111 figs., 24 tabs., 15 refs

  13. Preliminary report on the promise of accelerator-driven natural-uranium-fueled light-water-moderated breeding power reactors

    International Nuclear Information System (INIS)

    A new concept for a power breeder reactor that consists of an accelerator-driven subcritical thermal fission system is proposed. In this system an accelerator provides a high-energy proton beam which interacts with a heavy-element target to produce, via spallation reactions, an intense source of neutrons. This source then drives a natural-uranium-fueled, light-water-moderated and -cooled subcritical blanket which both breeds new fuel and generates heat that can be converted to electrical power. The report given presents a general layout of the resulting Accelerator Driven Light Water Reactor (ADLWR), evaluates its performance, discusses its fuel cycle characteristics, and identifies the potential contributions to the nuclear energy economy this type of power reactor might make. A light-water thermal fission system is found to provide an attractive feature when designed to be source-driven. The equilibrium fissile fuel content that gives the highest energy multiplication is approximately equal to the content of 235U in natural uranium. Consequently, natural-uranium-fueled ADLWRs that are designed to have the highest energy generation per source neutron are also fuel-self-sufficient; that is, their fissile fuel content remains constant with burnup. This feature allows the development of a nuclear energy system that is based on the most highly developed fission technology available (the light water reactor technology) and yet has a simple and safe fuel cycle. ADLWRs will breed on natural uranium, have no doubling time limitation, and be free from the need for uranium enrichment or for the separation of plutonium. It appears that ADLWRs could also be efficiently operated with thorium fuel cycles and with denatured fuel cycles

  14. Draft report on compilation of generic safety issues for light water reactor nuclear power plants

    International Nuclear Information System (INIS)

    A generally accepted approach to characterizing the safety concerns in nuclear power plants is to express them as safety issues which need to be resolved. When such safety issues are applicable to a generation of plants of a particular design or to a family of plants of similar design, they are termed generic safety issues. Examples of generic safety issues are those related to reactor vessel embrittlement, control rod insertion reliability or strainer clogging. The safety issues compiled in this document are based on broad international experience. This compilation is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. Refs

  15. Technical support for the EPRI [Electric Power Research Institute] debris coolability requirement for advanced light water reactors

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) advanced light water reactor (ALWR) design requirement for cavity sizing to promote long-term core debris coolability specifies that the reactor containment be arranged to afford a floor area for potential core debris spreading of at least 0.02 square meters per thermal megawatt of reactor power. This requirement has substantial experimental and analytical support as a sufficient basis to ensure long-term debris coolability. A logical framework and presentation of such support is provided in this document. The evidence as presented demonstrates that, following vessel failure, core debris in the cavity would consist of a mixture of fragments and a relatively continuous, but porous and cracked, phase that would be distributed relatively uniformly over the area that would be available per the requirement. Considering pertinent operative heat transfer phenomena, it is also demonstrated that the heat removal rate from core debris in this configuration would be sufficient to ensure debris coolability in the reactor cavity, even when the cavity is flooded after vessel failure. 29 refs., 6 figs., 8 tabs

  16. Safety Reassessment for Research Reactors in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant

    International Nuclear Information System (INIS)

    This publication aims to provide practical information on the performance of safety reassessment for research reactors. The information provided is relevant for safety reassessment for all types of nuclear installation in the light of the accident that occurred in March 2011 at the Fukushima Daiichi nuclear power plant in Japan following a severe off-shore earthquake and subsequent tsunami. Flooding of the power plant and damage to equipment due to the tsunami resulted in an extended station blackout, loss of core cooling, fuel melting, hydrogen explosions and releases of radioactive material to the surrounding region, with significant contamination of the environment. The available experience from this accident will be useful for defining and implementing measures to prevent the occurrence of any accident involving a large release of radioactive material at nuclear installations, including at a research reactor, in the future. The majority of research reactors were built to earlier safety standards, which are not fully consistent with the IAEA safety standards and with the defence in depth concept. In particular, for many research reactors the design of structures, systems and components important to safety is not in accordance with the criterion for common cause failure (i.e. the ability to withstand the failure of two or more structures, systems or components due to a single specific event or cause), and the confinement or containment buildings of several research reactors located near populated areas have deficiencies in their leaktightness. In addition, the safety analyses for many research reactors have not been updated to take into account modifications of the facilities and changes in the characteristics of their sites and site vicinity areas. These elements and the feedback from the Fukushima Daiichi accident justify a revision of the safety analysis for these facilities through the performance of a safety reassessment. This publication provides

  17. Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Izumi, Fumio

    1995-12-01

    The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author).

  18. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  19. Fundamentals of nuclear power plants with light water reactors. Pt. 1

    International Nuclear Information System (INIS)

    The authors give a comprehensive picture (in two parts) of modern LWR reactors. All technical constructive and physical details of BWR and PWR reactors are described and compared. The first part describes the different cooling systems and their components, including control systems. In the second part, the layout of the reactor core, fuel assemblies, instrumentation and thermohydraulic aspects are reported on. (TK)

  20. Identification and evaluation of facilitation techniques for decommissioning light water power reactors

    International Nuclear Information System (INIS)

    This report describes a study sponsored by the US Nuclear Regulatory Commission to identify practical techniques to facilitate the decommissioning of nuclear power generating facilities. The objective of these ''facilitation techniques'' is to reduce the radioactive exposures and/or volumes of waste generated during the decommissioning process. The report presents the possible facilitation techniques identified during the study and discusses the corresponding facilitation of the decommissioning process. Techniques are categorized by their applicability of being implemented during the three stages of power reactor life: design/construction, operation, or decommissioning. Detailed cost-benefit analyses were performed for each technique to determine the anticipated exposure and/or radioactive waste reduction; the estimated costs for implementing each technique were then calculated. Finally, these techniques were ranked by their effectiveness in facilitating the decommissioning process. This study is a part of the Nuclear Regulatory Commission's evaluation of decommissioning policy and its modification of regulations pertaining to the decommissioning process. The findings can be used by the utilities in the planning and establishment of activities to ensure that all objectives of decommissioning will be achieved

  1. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

  2. NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

    International Nuclear Information System (INIS)

    The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

  3. NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

    International Nuclear Information System (INIS)

    The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

  4. Light water reactor safety research project

    International Nuclear Information System (INIS)

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  5. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  6. Detailed description of an SSAC at the facility level for light water moderated (off-load refueled) power reactor facilities

    International Nuclear Information System (INIS)

    This report is intended to provide the technical details of an effective State Systems of Accounting for and Control of Nuclear Material (SSAC) which Member States may use, if they wish, to establish and maintain their SSACs. It is expected that systems designed along the lines described would be effective in meeting the objectives of both national and international systems for nuclear material accounting and control. This document accordingly provides a detailed description of a system for the accounting for and control of nuclear material in an off-load refueled light water moderated power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Progress Report for Year 1, Quarter 2 (January - March 2002)

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the 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 about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  9. Corrosion failures and related research subjects in light water reactor power plants

    International Nuclear Information System (INIS)

    The purpose of this paper is to review the recent corrosion incidents and the remedial actions taken by the utility industry. The following corrosion failures and related subjects are presented in this paper. (1) Intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs). (2) Interdendritic stress corrosion cracking (IDSCC) of stainless steel piping weldments in BWRs. (3) Irradiation-assisted stress corrosion cracking (IASCC) in in-core component of nonsensitized austenitic stainless steel. (4) Intergranular attack (IGA) of steam generator tube material in pressurized water reactors (PWRs). (5) Corrosion fatigue and erosion corrosion of carbon steel piping. (6) SCC of reactor component materials. (7) Microbial induced corrosion. (author)

  10. Light Water Reactor Sustainability Program Power Uprate Research and Development Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang

    2011-09-01

    The economic incentives for low-cost electricity generation will continue to drive more plant owners to identify safe and reliable methods to increase the electrical power output of the current nuclear power plant fleet. A power uprate enables a nuclear power plant to increase its electrical output with low cost. However, power uprates brought new challenges to plant owners and operators. These include equipment damage or degraded performance, and unanticipated responses to plant conditions, etc. These problems have arisen mainly from using dated design and safety analysis tools and insufficient understanding of the full implications of the proposed power uprate or from insufficient attention to detail during the design and implementation phase. It is essential to demonstrate that all required safety margins have been properly retained and the existing safety level has been maintained or even increased, with consideration of all the conditions and parameters that have an influence on plant safety. The impact of the power uprate on plant life management for long term operation is also an important issue. Significant capital investments are required to extend the lifetime of an aging nuclear power plant. Power uprates can help the plant owner to recover the investment costs. However, plant aging issues may be aggravated by the power uprate due to plant conditions. More rigorous analyses, inspections and monitoring systems are required.

  11. Generic safety issues for nuclear power plants with light water reactors and measures taken for their resolution

    International Nuclear Information System (INIS)

    The IAEA Conference on 'The Safety of Nuclear Power: Strategy for the Future' in 1991 was a milestone in nuclear safety. Two of the important items addressed by this conference were ensuring and enhancing safety of operating plants and treatment of nuclear power plants built to earlier safety standards. A number of publications related to these two items issued subsequent to this conference were: A Common Basis for Judging the Safety of Nuclear Power Plants Built to Earlier Standards, INSAG-9 (1995), the IAEA Safety Guide 50-SG-O12, periodic Safety Review of Operational Nuclear Power Plants (1994) and an IAEA publication on the Safety Evaluation of Operating Nuclear Power Plants Built to Earlier Standards - A Common Basis for Judgement (1997). Some of the findings of the 1991 Conference have not yet been fully addressed. An IAEA Symposium on reviewing the Safety of Existing Nuclear Power Plants in 1996 showed that there is an urgent need for operating organizations and national authorities to review operating nuclear power plants which do not meet the high safety levels of the vast majority of plants and to undertake improvements with assistance from the international community if required. Safety reviews of operating nuclear power plants take on added importance in the context of the Convention on Nuclear safety and its implementation. The purpose of this TECDOC compilation based on broad international experience, is to assist the Member States in the reassessment of operating plants by providing a list of generic safety issues identified in nuclear power plants together with measures taken to resolve these issues. These safety issues are generic in nature with regard to light water reactors and the measures for their resolution are for use as a reference for the safety reassessment of operating plants. The TECDOC covers issues thought to be significant to Member States based on consensus process. It provides an introduction to the use of generic safety issues for

  12. Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

    International Nuclear Information System (INIS)

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U233-Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  13. Light Water Reactor Sustainability Program: Computer-based procedure for field activities: results from three evaluations at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna [Idaho National Laboratory; Bly, Aaron [Idaho National Laboratory; LeBlanc, Katya [Idaho National Laboratory

    2014-09-01

    Nearly all activities that involve human interaction with the systems of a nuclear power plant are guided by procedures. The paper-based procedures (PBPs) currently used by industry have a demonstrated history of ensuring safety; however, improving procedure use could yield tremendous savings in increased efficiency and safety. One potential way to improve procedure-based activities is through the use of computer-based procedures (CBPs). Computer-based procedures provide the opportunity to incorporate context driven job aids, such as drawings, photos, just-in-time training, etc into CBP system. One obvious advantage of this capability is reducing the time spent tracking down the applicable documentation. Additionally, human performance tools can be integrated in the CBP system in such way that helps the worker focus on the task rather than the tools. Some tools can be completely incorporated into the CBP system, such as pre-job briefs, placekeeping, correct component verification, and peer checks. Other tools can be partly integrated in a fashion that reduces the time and labor required, such as concurrent and independent verification. Another benefit of CBPs compared to PBPs is dynamic procedure presentation. PBPs are static documents which limits the degree to which the information presented can be tailored to the task and conditions when the procedure is executed. The CBP system could be configured to display only the relevant steps based on operating mode, plant status, and the task at hand. A dynamic presentation of the procedure (also known as context-sensitive procedures) will guide the user down the path of relevant steps based on the current conditions. This feature will reduce the user’s workload and inherently reduce the risk of incorrectly marking a step as not applicable and the risk of incorrectly performing a step that should be marked as not applicable. As part of the Department of Energy’s (DOE) Light Water Reactors Sustainability Program

  14. Light Water Reactor Sustainability Program: Computer-based procedure for field activities: results from three evaluations at nuclear power plants

    International Nuclear Information System (INIS)

    Nearly all activities that involve human interaction with the systems of a nuclear power plant are guided by procedures. The paper-based procedures (PBPs) currently used by industry have a demonstrated history of ensuring safety; however, improving procedure use could yield tremendous savings in increased efficiency and safety. One potential way to improve procedure-based activities is through the use of computer-based procedures (CBPs). Computer-based procedures provide the opportunity to incorporate context driven job aids, such as drawings, photos, just-in-time training, etc into CBP system. One obvious advantage of this capability is reducing the time spent tracking down the applicable documentation. Additionally, human performance tools can be integrated in the CBP system in such way that helps the worker focus on the task rather than the tools. Some tools can be completely incorporated into the CBP system, such as pre-job briefs, placekeeping, correct component verification, and peer checks. Other tools can be partly integrated in a fashion that reduces the time and labor required, such as concurrent and independent verification. Another benefit of CBPs compared to PBPs is dynamic procedure presentation. PBPs are static documents which limits the degree to which the information presented can be tailored to the task and conditions when the procedure is executed. The CBP system could be configured to display only the relevant steps based on operating mode, plant status, and the task at hand. A dynamic presentation of the procedure (also known as context-sensitive procedures) will guide the user down the path of relevant steps based on the current conditions. This feature will reduce the user's workload and inherently reduce the risk of incorrectly marking a step as not applicable and the risk of incorrectly performing a step that should be marked as not applicable. As part of the Department of Energy's (DOE) Light Water Reactors Sustainability Program

  15. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

  16. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  17. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  18. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  19. An evaluation of light water breeder reactor system (LWBR) as an alternative for nuclear power generation in Brazil

    International Nuclear Information System (INIS)

    The LWBR system as an alternative for nuclear power generation in Brazil, was technically and economically evaluated. The LWBR system has been characterized comparatively with the Pressurized Water Reactors through technological and investment cost analysis and through the analysis of the processes and unit costs of the fuel cycle stages. The characteristics of the LWBR system in comparison to the PWR system, with respect to utilization and cumulative consumption of uranium and thorium resources, fuel cycle processes and associated costs have been determined for possible alternatives of nuclear power participation in the Brazilian hidro-thermal electricity generating system. The analysis concluded that the LWBR system does not represent an attractive alternative for nuclear power generation in Brazil and even has no potential to compete with conventional Pressurized Water Reactors. (Author)

  20. Development of a human reliability analysis procedure for a low power/shutdown probabilistic safety assessment in pressurized light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kang, D. I.; Sung, T. Y.; Park, J. H.; Kim, T. W.; Han, S. H.; Kim, K. Y.; Yang, J. E.; Jung, W. D.; Lee, Y. H.; Hwang, M. J.

    1997-09-01

    A human reliability analysis (HRA) procedure is developed for a low power/shutdown probalistic safety assessment (PSA) in pressurized light water reactors. At first, the HRA procedure developed is based on the two major current methods: THERP (technique for human error rate prediction) and SHARP (systematic human action reliability procedure). Then, it focuses on the specific situation of low power and shutdown operation of pressurized light water reactors. Major characteristics of the HRA procedure are as follows; 1) The use of the work sheet developed increase the plausibility and credibility of the quantification process of human actions and enable use to trace easily it. 2) The explicit use of decision tree could partly eliminate the possible subjectiveness in human reliability analyst`s judgement used for HRA. It is expected that the HRA procedure developed allow human reliability analyst to perform a systematic and consistent HRA. (author). 26 refs., 13 tabs., 8 figs.

  1. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  2. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 2. Appendices

    International Nuclear Information System (INIS)

    This volume contains the following appendices: LWR fuel cycle, handling and storage of spent fuel, termination case considerations (use of coal-fired power plants to replace nuclear plants), increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data, characteristics of nuclear fuel, away-from-reactor storage concept, spent fuel storage requirements for higher projected nuclear generating capacity, and physical protection requirements and hypothetical sabotage events in a spent fuel storage facility

  3. Light water reactor safety research

    International Nuclear Information System (INIS)

    As the technology of light water reactors (LWR) was being commercialized, the German Federal Government funded the reactor safety research program, which was conducted by national research centers, universities, and industry, and which led to the establishment, in early 1972, of the Nuclear Safety Project in Karlsruhe. In the seventies, the PNS project mainly studied the loss-of-coolant accident. Numerous experiments were run and computer codes developed for this purpose. In the eighties, the Karlsruhe Nuclear Research Center contributed to the German Risk Study, investigating especially core meltdown accidents under the impact of the events at Three Mile Island-2 and Chernobyl-4. Safety research in the nineties is concentrated on the requirements of future reactor generations, such as the European Pressurized Water Reactor (EPR) or potential approaches which, at the present time, are discernible only as tentative theoretical designs. (orig.)

  4. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  5. Power calibrations for TRIGA reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  6. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 1. Executive summary and text

    International Nuclear Information System (INIS)

    The Generic Environmental Impact Statement on spent fuel storage was prepared by the Nuclear Regulatory Commission staff in response to a directive from the Commissioners published in the Federal Register, September 16, 1975 (40 FR 42801). The Commission directed the staff to analyze alternatives for the handling and storage of spent light water power reactor fuel with particular emphasis on developing long range policy. Accordingly, the scope of this statement examines alternative methods of spent fuel storage as well as the possible restriction or termination of the generation of spent fuel through nuclear power plant shutdown. Volume 1 includes the executive summary and the text

  7. Power Reactors. Appendix VIII

    International Nuclear Information System (INIS)

    Decommissioning of nuclear facilities in many countries has evolved into a mature industry that has benefited from experience gained from previous projects and decommissioning costs can now be estimated to a good degree of accuracy. As a result of lessons learned, future decommissioning projects can be performed with higher levels of efficiency. Decommissioning of old power reactors is in progress in several countries. In some cases, decommissioning has been completed (i.e. plant sites have been released from regulatory control), while in other countries decommissioning is still in progress. Several large power reactors have been successfully decommissioned since 1995. The key areas of particular importance for decommissioning are decontamination, radiation protection, dismantling and demolition. The technologies which can be used for these tasks are commonly available on the market, but effective decommissioning still depends on an optimal choice of technologies, including site specific developments. It is not possible to recommend the use of a single specific technology for dismantling, demolition, segmentation or decontamination; rather, it is good practice to take into account as much information as possible from other decommissioning projects and to draw comparisons between various techniques in order to choose the one with the best performance in a particular situation. The exchange of information on all types of decommissioning experience, including decommissioning techniques and their applicability as well as disadvantages for specific tasks, is taking place on various levels, such as: — Collaborative working groups established by international organizations such as the IAEA, the OECD Nuclear Energy Agency and the European Commission and the publication of technical reports by such organizations; — National and international conferences; — Bilateral or multilateral cooperation and information exchange between organizations with responsibilities for

  8. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided 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.

  9. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided 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

  10. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided 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

  11. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided 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.

  12. Inherently safe light water reactors

    International Nuclear Information System (INIS)

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

  13. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  14. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the 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 about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

  15. Fractals in Power Reactor Noise

    International Nuclear Information System (INIS)

    In this work the non- lineal dynamic problem of power reactor is analyzed using classic concepts of fractal analysis as: attractors, Hausdorff-Besikovics dimension, phase space, etc. A new non-linear problem is also analyzed: the discrimination of chaotic signals from random neutron noise signals and processing for diagnosis purposes. The advantages of a fractal analysis approach in the power reactor noise are commented in details

  16. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

  17. Advanced light water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Giedraityte, Zivile [Helsinki University of Technology, Otaranta 8D-84, 02150 Espoo (Finland)

    2008-07-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  18. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

  19. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U3O8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  20. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, walk-away safe design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (O and M) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

  1. Experimental power reactor

    International Nuclear Information System (INIS)

    The following five topics are discussed using figures and diagrams: (1) energy storage and transfer program, (2) thermomechanical analysis, (3) a steam dual-cycle power conversion system for the EPR, (4) EPR tritium facility scoping studies, and (5) vacuum systems

  2. Neutronics of nuclear power reactors

    International Nuclear Information System (INIS)

    This review, prepared on the occasion of 25th ETAN Conference describes the research activities in the field of neutronics which started in 1947. A number of researchers in Yugoslav Institutes was engaged in development of neutronics theory and calculation methods related to power reactors since 1960. To illustrate the activities of Yugoslav authors, this review contains the list of the most important relevant papers published in international journals

  3. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  4. Lifetime and ageing management of nuclear power plants - a brief overview of some light-water reactor component ageing degradation problems and ways of mitigation

    International Nuclear Information System (INIS)

    The susceptibility and tendency for nuclear power plant (NPP) components to undergo changes in their mechanical and physical properties in the course of the NPP lifetime is generally termed as ''ageing''. The light water reactor (LWR) nuclear environment is not a benign one; it is characterized not only by fast neutrons, which can degrade metallic structures, but also relatively high temperatures of coolant water (around 300oC) which may contain impurities (e.g. sulphate, chloride, peroxides) which can cause corrosion. The hydrodynamical conditions of the coolant in steam generators, for example, can cause thermal fatigue, erosion and corrosion; suspended particulate matter can accelerate erosion processes which can lead to wall thinning and lowering of safety margins. The paper describes a few ageing phenomena which can be considered important due to their impact on nuclear safety issues; their influence on economic aspects (availability) of NPP is outlined. Mitigation measures for NPP component ageing problems are given together with recommendations for addressing such issues in NPP now and in the future. (author)

  5. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  6. Analysis of higher power research reactors' parameters

    International Nuclear Information System (INIS)

    The objective of this monograph was to analyze and compare parameters of different types of research reactors having higher power. This analysis could be used for decision making and choice of a reactor which could possibly replace the existing ageing RA reactor in Vinca. Present experimental and irradiation needs are taken into account together with the existing reactors operated in our country, RB and TRIGA reactor

  7. Roadmap for implementation of light water reactor decommissioning

    International Nuclear Information System (INIS)

    While decommissioning of Tokai-mura reactor and JATR reactor has already started in Japan, Tsuruga reactor is announced shutdown in 2010 as the first decommissioning of commercial light water reactor (LWR). In 2030s or may be more earlier due to economic reasons, decommissioning of LWRs will take place in succession. Since rational decommissioning needs operating data of individual plants, ample time should be allowed for planning the reactor decommissioning. Committee of Nuclear Power Engineering Cooperation (NUPEC) had identified relevant issues to implement LWR decommissioning and established roadmaps showing fundamental approaches to solve seventeen items categorized in seven areas as action items. Harmonization of policy, regulations and technology development as a whole and reflection of accumulated lessons learned from overseas decommissioning experiences needed further study. (T. Tanaka)

  8. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  9. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  10. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  11. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  12. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  13. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  14. Performance indicators for power reactors

    International Nuclear Information System (INIS)

    A review of Canadian and worldwide performance indicator definitions and data was performed to identify a set of indicators that could be used for comparison of performance among nuclear power plants. The results of this review are to be used as input to an AECB team developing a consistent set of performance indicators for measuring Canadian power reactor safety performance. To support the identification of performance indicators, a set of criteria was developed to assess the effectiveness of each indicator for meaningful comparison of performance information. The project identified a recommended set of performance indicators that could be used by AECB staff to compare the performance of Canadian nuclear power plants among themselves, and with international performance. The basis for selection of the recommended set and exclusion of others is provided. This report provides definitions and calculation methods for each recommended performance indicator. In addition, a spreadsheet has been developed for comparison and trending for the recommended set of indicators. Example trend graphs are included to demonstrate the use of the spreadsheet. (author). 50 refs., 11 tabs., 3 figs

  15. Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

    Directory of Open Access Journals (Sweden)

    Geoff Recktenwald

    2013-04-01

    Full Text Available Zirconium dioxide has received particular attention as a fuel matrix because of its ability to form a solid solution with transuranic elements, natural radiation stability and desirable mechanical properties. However, zirconium dioxide has a lower coefficient of thermal conductivity than uranium dioxide and this presents an obstacle to the deployment of these fuels in commercial reactors. Here we show that axial doping of a zirconium dioxide based fuel with erbium reduces power peaking and fuel temperature. Full core simulations of a modified AP1000 core were done using MCNPX 2.7.0. The inert matrix fuel contained 15 w/o transuranics at its beginning of life and constituted 28% of the assemblies in the core. Axial doping reduced power peaking at startup by more than ~23% in the axial direction and reduced the peak to average power within the core from 1.80 to 1.44. The core was able to remain critical between refueling while running at a simulated 2000 MWth on an 18 month refueling cycle. The results show that the reactor would maintain negative core average reactivity and void coefficients during operation. This type of fuel cycle would reduce the overall production of transuranics in a pressurized water reactor by 86%.

  16. Critical analysis of accident scenario and consequences modelling applied to light-water reactor power plants for accident categories beyond the design basis accident (DBA)

    International Nuclear Information System (INIS)

    A critical analysis and sensitivity study of the modelling of accident scenarios and environmental consequences are presented, for light-water reactor accident categories beyond the standard design-basis-accident category. The first chapter, on ''source term'' deals with the release of fission products from a damaged core inventory and their migration within the primary circuit and the reactor containment. Particular attention is given to the influence of engineering safeguards intervention and of the chemical forms of the released fission products. The second chapter deals with their release to the atmosphere, transport and wet or dry deposition, outlining relevant partial effects and confronting short-duration or prolonged releases. The third chapter presents a variability analysis, for environmental contamination levels, for two extreme hypothetical scenarios, evidencing the importance of plume rise. A numerical plume rise model is outlined

  17. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Science.gov (United States)

    2010-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which...

  18. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  19. Wind Powering America: The Next Steps in North Carolina

    Energy Technology Data Exchange (ETDEWEB)

    Banks, Jennifer L. [North Carolina Solar Center; Scanlin, Dennis [Appalachian State University; Quinlan, Paul [North Carolina Sustainable Energy Association

    2013-06-18

    The goal of this project is to apply the WPA’s proactive outreach strategy to the problem of educating the public about the likely transmission infrastructure developments concomitant to the significant development of wind energy resources in North Carolina. Given the lead time to develop significant new transmission infrastructure (5-10 years), it is critical to begin this outreach work today, so that wind resources can be developed to adequately meet the 20% by 2030 goal in the mid- to long-term (10-20 years). The project team planned to develop a transmission infrastructure outreach campaign for North Carolina by: (1) convening a utility interest group (UIG) of the North Carolina Wind Working Group (NC WWG) consisting of electric utilities in the state and the Southeast; and (2) expanding outreach to local and state government officials in North Carolina.

  20. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  1. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  2. Small size modular fast reactors in large scale nuclear power

    International Nuclear Information System (INIS)

    The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

  3. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  4. Higher power density TRIGA research reactors

    International Nuclear Information System (INIS)

    The uranium zirconium hydride (U-ZrH) fuel is the fundamental feature of the TRIGA family of reactors that accounts for its widely recognized safety, good performance, economy of operation, and its acceptance worldwide. Of the 65 TRIGA reactors or TRIGA fueled reactors, several are located in hospitals or hospital complexes and in buildings that house university classrooms. These examples are a tribute to the high degree of safety of the operating TRIGA reactor. In the early days, the majority of the TRIGA reactors had power levels in the range from 10 to 250 kW, many with pulsing capability. An additional number had power levels up to 1 MW. By the late 1970's, seven TRIGA reactors with power levels up to 2 MW had been installed. A reduction in the rate of worldwide construction of new research reactors set in during the mid 1970's but construction of occasional research reactors has continued until the present. Performance of higher power TRIGA reactors are presented as well as the operation of higher power density reactor cores. The extremely safe TRIGA fuel, including the more recent TRIGA LEU fuel, offers a wide range of possible reactor configurations. A long core life is assured through the use of a burnable poison in the TRIGA LEU fuel. In those instances where large neutron fluxes are desired but relatively low power levels are also desired, the 19-rod hexagonal array of small diameter fuel rods offers exciting possibilities. The small diameter fuel rods have provided extremely long and trouble-free operation in the Romanian 14 MW TRIGA reactor

  5. Power source device for reactor recycling pump

    International Nuclear Information System (INIS)

    The device of the present invention prevents occurrence of an accident of a reactor forecast upon spontaneous power stoppage, loss of power source or trip of the reactor. Namely, a AC/DC converter and a DC/AC connector having an AC voltage frequency controller are connected in series between an AC (bus) in the plant and reactor recycling pumps. A DC voltage controller, a superconductive energy storing device and an excitation power source are connected to the input of the DC/AC converter. The control device receives signals of the spontaneous power stoppage, loss of power source or trip of the reactor to maintain the output voltage of the superconductive energy storing device to a predetermined value. Further, the ratio of AC power voltage and the frequency of AC voltage to be supplied to the reactor recycling pumps is constantly varied to control the flow rate of the pump to a predetermined value. With such procedures, a power source device for the reactor recycling pumps compact in size, easy for maintenance and having high reliability can be realized by adopting a static-type superconductive energy storing device as an auxiliary power source for the reactor recycling pumps. (I.S.)

  6. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  7. Light Water Reactor Sustainability Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  8. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

  9. Isotopic signature of atmospheric xenon released from light water reactors

    International Nuclear Information System (INIS)

    A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of 135Xe, 133mXe, 133Xe and 131mXe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios

  10. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  11. Risks of nuclear energy technology safety concepts of light water reactors

    CERN Document Server

    Kessler, Günter; Schlüter, Franz-Hermann

    2014-01-01

    The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on?reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: ? A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Ch

  12. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  13. Modularity of the MIT Pebble Bed Reactor for use by the commercial power industry

    OpenAIRE

    Hanlon-Hyssong, Jaime E.

    2008-01-01

    CIVINS The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and the US. To make smaller 120 Mwe reactors economically competitive with larger 1500 Mwe traditional light water reactors changes in the way these plants are built are needed. Economies of production need to be sufficiently large to compete with economies of sca...

  14. Conceptual design of light ion beam inertia nuclear fusion reactors

    International Nuclear Information System (INIS)

    Light ion beam, inertia nuclear fusion system drew attention recently as one of the nuclear fusion systems for power reactors in the history of the research on nuclear fusion. Its beginning seemed to be the judgement that the implosion of fusion fuel pellets with light ions can be realized with the light ions which can be obtained in view of accelerator techniques. Of course, in order to generate practically usable nuclear fusion reaction by this system and maintain it, many technical difficulties must be overcome. This research was carried out for the purpose of discovering such technical problems and searching for their solution. At the time of doing the works, the following policy was adopted. Though their is the difference of fine and rough, the design of a whole reactor system is performed conformably. In order to make comparison with other reactor types and nuclear fusion systems, the design is carried out as the power plant of about one million kWe output. As the extent of the design, the works at conceptual design stage are performed to present the concept of design which satisfies the required function. Basically, the design is made from conservative standpoint. This research of design was started in 1981, and in fiscal 1982, the mutual adjustment among the design of respective parts was performed on the basis of the results in 1981, and the possible revision and new proposal were investigated. (Kako, I.)

  15. Thermionic reactors for space nuclear power

    Science.gov (United States)

    Griaznov, Georgii M.; Zhabotinskii, Evgenii E.; Serbin, Victor I.; Zrodnikov, Anatolii V.; Pupko, Victor Ia.; Ponomarev-Stepnoi, Nikolai N.; Usov, V. A.; Nikolaev, Iu. V.

    Compact thermionic nuclear reactor systems with satisfactory mass performance are competitive with space nuclear power systems based on the organic Rankine and closed Brayton cycles. The mass characteristics of the thermionic space nuclear power system are better than that of the solar power system for power levels beyond about 10 kWe. Longlife thermionic fuel element requirements, including their optimal dimensions, and common requirements for the in-core thermionic reactor design are formulated. Thermal and fast in-core thermionic reactors are considered and the ranges of their sensible use are discussed. Some design features of the fast in-core thermionic reactors cores (power range to 1 MWe) including a choice of coolants are discussed. Mass and dimensional performance for thermionic nuclear power reactor system are assessed. It is concluded that thermionic space nuclear power systems are promising power supplies for spacecrafts and that a single basic type of thermionic fuel element may be used for power requirements ranging to several hundred kWe.

  16. MIT research reactor. Power uprate and utilization

    International Nuclear Information System (INIS)

    The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

  17. Controlling hydrogen behavior in light water reactors

    International Nuclear Information System (INIS)

    In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

  18. Is light water reactor technology sustainable?

    International Nuclear Information System (INIS)

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

  19. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  20. New generation of reactors for space power

    International Nuclear Information System (INIS)

    Space nuclear reactor power is expected to enable many new space missions that will require several times to several orders of magnitude anything flown in space to date. Power in the 100-kW range may be required in high earth orbit spacecraft and planetary exploration. The technology for this power system range is under development for the Department of Energy with the Los Alamos National Laboratory responsible for the critical components in the nuclear subsystem. The baseline design for this particular nuclear sybsystem technology is described in this paper; additionally, reactor technology is reviewed from previous space power programs, a preliminary assessment is made of technology candidates covering an extended power spectrum, and the status is given of other reactor technologies

  1. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  2. Safety aspects of designs for future light water reactors (evolutionary reactors)

    International Nuclear Information System (INIS)

    The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

  3. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  4. Thorium utilization in power reactors

    International Nuclear Information System (INIS)

    In this work the recent (prior to Aug, 1976) literature on thorium utilization is reviewed briefly and the available information is updated. After reviewing the nuclear properties relevant to the thorium fuel cycle we describe briefly the reactor systems that have been proposed using thorium as a fertile material. (author)

  5. Powering light boats

    Energy Technology Data Exchange (ETDEWEB)

    Affolter, J.-F.

    1999-07-01

    This short paper reports on a workshop held in Horw, Switzerland, on the application potential of low-temperature fuel cells. It looks at the increasing interest in small, electrically-driven boats in Europe and, in particular, at the option of powering them with fuel cells. A prototype implementation of a test boat is described. Powered by a 300-watt portable PEFC fuel cell constructed by the Paul Scherrer Institute (PSI) and the University of Applied Science in Solothurn, Switzerland, the boat is part of a project to study the technical feasibility of such drive systems for leisure craft and small boats. The paper addresses hydrogen security questions, energy-efficient drives and controllers, maintenance questions and discusses if a combination of fuel cells and photovoltaics would be viable. Also, the design of the boat's hull and the propellers that are used are looked at critically. The economic viability of fuel-cell powered boats is examined. The authors state that they are convinced that such fuel cell drives could be commercially interesting for larger passenger boats, too.

  6. Environmentally assisted cracking in Light Water Reactors

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  7. Heat pipe reactors for space power applications

    Science.gov (United States)

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

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  8. The Swedish Zero Power Reactor R0

    Energy Technology Data Exchange (ETDEWEB)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-15

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

  9. Status of advanced light water reactor designs 2004

    International Nuclear Information System (INIS)

    The report is intended to be a source of reference information for interested organizations and individuals. Among them are decision makers of countries considering implementation of nuclear power programmes. Further, the report is addressed to government officials with an appropriate technical background and to research institutes of countries with existing nuclear programmes that wish to be informed on the global status in order to plan their nuclear power programmes including both research and development efforts and means for meeting future. The future utilization of nuclear power worldwide depends primarily on the ability of the nuclear community to further improve the economic competitiveness of nuclear power plants while meeting stringent safety requirements. The IAEA's activities in nuclear power technology development include the preparation of status reports on advanced reactor designs to provide all interested IAEA Member States with balanced and objective information on advances in nuclear plant technology. In the field of light water reactors, the last status report published by the IAEA was 'Status of Advanced Light Water Cooled Reactor Designs: 1996' (IAEA-TECDOC-968). Since its publication, quite a lot has happened: some designs have been taken into commercial operation, others have achieved significant steps toward becoming commercial products, including certification from regulatory authorities, some are in a design optimization phase to reduce capital costs, development for other designs began after 1996, and a few designs are no longer pursued by their promoters. With this general progress in mind, on the advice and with the support of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for Light Water Reactors (LWRs), the IAEA has prepared this new status report on advanced LWR designs that updates IAEA-TECDOC-968, presenting the various advanced LWR designs in a balanced way according to a common outline

  10. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

  11. Power for all? Electricity and uneven development in North Carolina

    Science.gov (United States)

    Harrison, Conor M.

    Many towns in eastern North Carolina face a number of challenges common to the rural South, including high rates of poverty and diminishing employment opportunities. However, some residents of this region also confront a unique hardship---electricity prices that are vastly higher than those of surrounding areas. This dissertation examines the origins of pricing inequalities in the electricity market of eastern North Carolina---namely how such inequalities developed and their role in the production of racial and economic disparities in the South. This dissertation examines the evolving relations between federal and state agencies, corporations, and electric utilities, and asks why these interactions produced varying social outcomes across different places and spatial settings. The research focuses on the origins and subsequent development of electric utilities in eastern North Carolina, and examines how electricity as a material technology interacted with geographies of race and class, as well as the dictates of capital accumulation. This approach enables a rethinking of several concepts that are rarely examined by scholars of electric utilities, most notably the monopoly service territory, which I argue served as a spatial fix to accumulation problems in the industry. Further, examining the way that electric utilities developed in North Carolina during the 20th century brings to the forefront the at times contradictory relationships among systems of electricity provision, Jim Crow segregation, the Progressive Era, and the New Deal. Such a focus highlights the important role that the control of electricity provision played in shaping racial inequalities that continue to persist in the region. With most urban areas were electrified in the 1930s, the research also traces the electricity distribution lines as they moved out of cities through rural electrification programs, a shift that highlights the state as a multi-scalar and variegated actor that both aided and

  12. New reactor technology: safety improvements in nuclear power systems.

    Science.gov (United States)

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  13. New reactor technology: safety improvements in nuclear power systems.

    Science.gov (United States)

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  14. High power cladding light strippers

    Science.gov (United States)

    Wetter, Alexandre; Faucher, Mathieu; Sévigny, Benoit

    2008-02-01

    The ability to strip cladding light from double clad fiber (DCF) fibers is required for many different reasons, one example is to strip unwanted cladding light in fiber lasers and amplifiers. When removing residual pump light for example, this light is characterized by a large numerical aperture distribution and can reach power levels into the hundreds of watts. By locally changing the numerical aperture (N.A.) of the light to be stripped, it is possible to achieve significant attenuation even for the low N.A. rays such as escaped core modes in the same device. In order to test the power-handling capability of this device, one hundred watts of pump and signal light is launched from a tapered fusedbundle (TFB) 6+1x1 combiner into a high power-cladding stripper. In this case, the fiber used in the cladding stripper and the output fiber of the TFB was a 20/400 0.06/0.46 N.A. double clad fiber. Attenuation of over 20dB in the cladding was measured without signal loss. By spreading out the heat load generated by the unwanted light that is stripped, the package remained safely below the maximum operating temperature internally and externally. This is achieved by uniformly stripping the energy along the length of the fiber within the stripper. Different adhesive and heat sinking techniques are used to achieve this uniform removal of the light. This suggests that these cladding strippers can be used to strip hundreds of watts of light in high power fiber lasers and amplifiers.

  15. General features of direct-cycle, supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR

  16. Liquid Metal Cooled Reactor for Space Power

    Science.gov (United States)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  17. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2005-01-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was

  18. Environmentally assisted cracking in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  19. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  20. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  1. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  2. Light water reactor lower head failure analysis

    International Nuclear Information System (INIS)

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response

  3. Gas-core reactor power transient analysis.

    Science.gov (United States)

    Kascak, A. F.

    1972-01-01

    The nuclear fuel in the gas-core reactor concept is a ball of uranium plasma radiating thermal photons. The photons are met by an inflowing hydrogen stream, which is seeded with submicron size, depleted uranium particles. A 'wall-burnout' condition exists if the thermal photons can reach the cavity liner because of insufficient absorption by the hydrogen. An analysis was conducted in order to determine the time for which the maximum steady state reactor power could be exceeded without damage to the cavity liner due to burnout. Wall-burnout time as a function of the power increase above the initial steady state condition is shown in a graph.

  4. Power Reactors in Small Packages

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R.

    1968-10-01

    This booklet discusses the introduction of nuclear power to remote places on earth where the resources of civilization are almost scarce. It also discusses nuclear power plants designed for use when warfare or natural catastrophes have wiped out the usual sources of energy, and in places beyond the reach of oil pipelines and coal trains. It also discusses how nuclear power may one day be used to manufacture chemical fuels for the world's vehicles when fossil fuels begin to run out.

  5. The safety of light water reactors

    International Nuclear Information System (INIS)

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  6. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in a previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes were considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the ''in-situ'' replacement of first walls using atomic coating processes were considered. The vapor deposition of carbon was shown to be promising

  7. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in the previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes was considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the in-situ replacement of first walls using atomic coating processes was considered. The vapor deposition of carbon was shown to be promising

  8. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  9. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  10. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  11. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  12. Utilization of thorium in power reactors

    International Nuclear Information System (INIS)

    The IAEA convened a Panel on the utilization of thorium in power reactors from 14 to 18 June 1965. 45 scientists from 14 countries and two international organizations took part in it. The proceedings of the Panel include 23 survey papers and brief reviews which stress the importance of utilizing thorium. A separate abstract was prepared for each of these papers. Refs, tabs, figs

  13. Simulation of power excursions - Osiris reactor

    International Nuclear Information System (INIS)

    Following the experimental work accomplished in the U.S.A. on Borax 1 and SPERT 1 and the accident of SL 1, the 'Commissariat a l'Energie Atomique' started a research program about the safety of its own swimming Pool reactors, with regard to power excursions. The first research work led to the design of programmed explosive charges, adapted to the simulation of a power excursion. This report describes the application of these methods to the investigation of Osiris safety. (author)

  14. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    Energy Technology Data Exchange (ETDEWEB)

    Seniuk, P.J.

    1996-12-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

  15. Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

  16. Technologies for Upgrading Light Water Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  17. Job-related doses in light water reactors

    International Nuclear Information System (INIS)

    The Treaty of 1957 establishing the European Atomic Energy Community, (EURATOM) was an essential prerequisite for the development of a strong nuclear industry in Europe. Among other things the Treaty provides that the Community shall lay down Basic Safety Standards for the protection of the health of workers and the general public against the dangers arising from ionizing radiation and ensure that they are applied. Following adoption of the Council Directive of 1980, the European Commission defined the basic principles of Justification, Optimization and Limitation to be applied in order to ensure the greatest possible protection of workers and the general public. Subsequently the Commission took initiatives in order to find ways of implementing these three basic principles in practical radiation protection. In 1980 the Commission in close collaboration with the leading nuclear power station operators, set up its own system of 'occupational radiation dose statistics from light water reactors operating in Western Europe'. This was designed for PWRs and BWRs, and the Commission benefited from the experience of neighbouring non-EC countries such as Sweden, Finland, Switzerland and Spain (not yet a member) operating nuclear power stations made by different manufacturers. The paper provides some general information on developments and trends in collective and individual doses to workers in nuclear power stations, based on a unique European databank of approximately 1000 operating reactor years. 9 figs

  18. Light water reactor piping system damping

    International Nuclear Information System (INIS)

    In this paper, based on a detailed evaluation and screening of existing damping data, a set of damping values are recommended for light water reactor piping systems. A multivariate regression model was used to identify the significant physical and response characteristics of piping systems. Although initially several experimental biases were identified that help explain the large variability in the existing data, these were ignored and only physical attributes were considered for the final recommendations. Of these twenty-two initial variables, only six were identified as being important to energy dissipation. Since the existing data is incomplete for certain variables, the identified parameters are not an exhaustive set. A regression analysis can only identify those parameters as significant that have a sufficient number and a wide spectrum of data points. Making several conservation assumptions, the six variable damping prediction equation was reduced to a damping table with two parameters: Response Level and Diameter. Pipe diameter is a convenient simple characteristic to represent system stiffness and hence support/pipe interaction, which tends to be a significant source of energy dissipation in piping systems

  19. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power Reactors.'' This guide describes a method NRC... decommissioning process for nuclear power reactors. The revision takes advantage of the 13 years...

  20. Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee`s annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs.

  1. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  2. Modular stellarator reactor: a fusion power plant

    International Nuclear Information System (INIS)

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment

  3. Modular stellarator reactor: a fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  4. Recent activities at the zero-power teaching reactor CROCUS

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.; Chawla, R., E-mail: gaetan.girardin@epfl.ch [Swiss Polytechnical School of Lausanne, Lausanne (Switzerland)

    2011-07-01

    CROCUS is a zero-power critical facility used mainly for educational purposes at the Swiss Federal Institute of Technology (EPFL) in Lausanne, Switzerland. It is a low-enriched-uranium fuelled, light-water moderated reactor, with the fission power limited to 100 W. The presentation will discuss the crucial role of CROCUS in teaching -- both as framework for reactor practicals offered to physics students at EPFL and as key educational tool in the recently established Swiss Master of Science in Nuclear Engineering. Regular development work is needed for the various instruments and components associated with the facility. As illustration, the recently completed refurbishment of the control rod system and the related calibration experiments will also be discussed.

  5. Compact approach to fusion power reactors

    International Nuclear Information System (INIS)

    The potential of the Reversed-Field Pinch (RFP) for development into an efficient, compact, copper-coil fusion reactor has been quantified by comprehensive parametric tradeoff studies. These compact systems promise to be competitive in size, power density, and cost to alternative energy sources. Conceptual engineering designs that largely substantiate these promising results have since been completed. This 1000-MWe(net) design is described along with a detailed rationale and physics/technology assessment for the compact approach to fusion

  6. Small and medium power reactors 1985

    International Nuclear Information System (INIS)

    This report is intended for designers and planners concerned with Small and Medium Power Reactors. It provides a record of the presentations during the meetings held on this subject at the Agency's General Conference in September 1985. This information should be useful as it indicates the principal findings and main conclusions and recommendations resulting from these meetings. A separate abstract was prepared for each of the 10 presentations in this report

  7. Utility Leadership in Defining Requirements for Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    It is appropriate, based on twenty five years of operating experience, that utilities take a position of leadership in developing the technical design and performance requirements for the next generations of nuclear electric generating plants. The U. S. utilities, through the Electric Power Research Institute, began an initiative in 1985 to develop such Utility requirements. Many international Utility organizations, including Korea Electric Power Corporation, have joined as full participants in this important Utility industry initiative. In light of the closer linkage among countries of the world due to rapid travel and telecommunications, it is also appropriate that there be international dialogue and agreement on the principal standards for nuclear power plant acceptability and performance. The Utility/EPRI Advanced Light Water Reactor Program guided by the ALRR Utility Steering Committee has been very successful in developing these Utility requirements. This paper will summarize the state of development of the ALRR Utility Requirements for Evolutionary Plants, recent developments in their review by the U. S. Nuclear Regulatory Commission, resolution of open issues, and the extension of this effort to develop a companion set of ALRR Utility Requirements for plants employing passive safety features

  8. Proposed Georgia-Alabama-South Carolina system power marketing policy and subsequent contracts

    International Nuclear Information System (INIS)

    This is an Environmental Assessment (Assessment) (DOE/EA-0935) evaluating the Power Marketing Policy and Subsequent Contracts between Southeastern and its customers. The Assessment evaluates two alternatives and the no action alternative. The proposed action is to market the power and energy available in the Georgia-Alabama-South Carolina System during the next ten years, with new power sales contracts of ten-year durations, to the customers set forth in Appendix A of the Assessment. In addition to the proposed alternative, the Assessment evaluates the alternative of extending existing contracts under the current marketing policy

  9. Overview of environmental materials degradation in light-water reactors

    International Nuclear Information System (INIS)

    This report provides a brief overview of analyses and conclusions reported in published literature regarding environmentally induced degradation of materials in operating light-water reactors. It is intended to provide a synopsis of subjects of concern rather than to address a licensing basis for any newly discovered problems related to reactor materials

  10. Nuclear power reactors and hydrogen storage systems

    International Nuclear Information System (INIS)

    Among conclusions and results come by, a nuclear-electric-hydrogen integrated power system was suggested as a way to prevent the energy crisis. It was shown that the hydrogen power system using nuclear power as a leading energy resource would hold an advantage in the current international situation as well as for the long-term future. Results reported provide designers of integrated nuclear-electric-hydrogen systems with computation models and routines which will allow them to explore the optimal solution in coupling power reactors to hydrogen producing systems, taking into account the specific characters of hydrogen storage systems. The models were meant for average computers of a type easily available in developing countries. (author)

  11. Improved once-through fuel cycles for light water reactors

    International Nuclear Information System (INIS)

    This paper is being presented at this time to provide preliminary technical and economic data to INFCE for use in comparisons of alternate nuclear systems. Programs to develop improved once-through fuel cycles for the light water reactor are under way in the United States; therefore, the information presented in this report is preliminary and will be updated in the future as it becomes available. In the meantime, the following limitations should be recognized when using the information in this report: 1. The paper quantifies fuel utilization improvements which should be technically feasible in reactors now operating or under construction and indicates the approximate time frame when the necessary development and demonstration could be completed. It does not attempt to estimate the rate at which these improvements would attain acceptance and use by the industry. 2. One particular set of PWR and one particular set of BWR nuclear reactor and fuel design characteristics are used as base cases, from which many of the improvements are estimated. Many plants operating and being built throughout the world of course differ in design features, fuel management schemes, and fuel utilization efficiencies from the base cases used in this paper. The degree of improvement obtainable in these other designs, for each type of change considered, will vary with each design. 3. The changes emphasized here could all be backfitted in existing plants. Other possible improvements are limited by the need to avoid reducing the power output or capacity factor of the plants. New plants could be designed to accommodate such changes without reducing the power output or capacity factor. This could yield greater improvement in fuel utilization than can be obtained in existing plants. This longer range potential has not been examined here

  12. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  13. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  14. Calculation of the optimum fuel distribution which maximizes the power output of a reactor

    International Nuclear Information System (INIS)

    Using optimal control techniques, the optimum fuel distribution - which maximizes the power output of a thermal reactor - is obtained. The nuclear reactor is described by a diffusion theory model with four energy groups and by assuming plane geometry. Since the analytical solution is impracticable, by using a perturbation method, a FORTRAN program was written, in order to obtain the numerical solution. Numerical results, for a thermal reactor light water moderated, non reflected, are shown. The fissile fuel material considered is Uranium-235. (Author)

  15. Application of fully ceramic microencapsulated fuels in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, C.; George, N.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States); Godfrey, A.; Terrani, K.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01

    This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

  16. Application of Fully Ceramic Microencapsulated Fuels in Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL

    2012-01-01

    This study aims to perform a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in Light Water Reactors (LWRs). In particular pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor. Using uranium-based fuel and transuranic (TRU) based fuel in TRistructural ISOtropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher physical density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the TRU based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, feasibility of core designs fully loaded with TRU FCM lattices was demonstrated using the NESTLE three-dimensional core simulator.

  17. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  18. Environmentally assisted cracking in light water reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  19. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  20. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  1. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  2. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    Science.gov (United States)

    Loeb, E.

    1961-01-17

    A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

  3. The risks of nuclear energy technology. Safety concepts of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Kern- und Energietechnk (IKET); Kessler, Guenter; Veser, Anke; Schlueter, Franz-Hermann

    2014-11-01

    Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

  4. The risks of nuclear energy technology. Safety concepts of light water reactors

    International Nuclear Information System (INIS)

    Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

  5. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

  6. Revised accident source terms for light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Soffer, L. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  7. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, Kevin L.; Ramuhalli, Pradeep; Brenchley, David L.; Coble, Jamie B.; Hashemian, Hash; Konnik, Robert; Ray, Sheila

    2012-09-14

    The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), NDE instrumentation development, universities, commercial NDE services and cable manufacturers, and Electric Power Research Institute (EPRI). The motivation for the R&D roadmap comes from the need to address the aging management of in-containment cables at nuclear power plants (NPPs).

  8. Cascade: a high-efficiency ICF power reactor

    International Nuclear Information System (INIS)

    Cascade attains a net power-plant efficiency of 49% and its cost is competitive with high-temperature gas-cooled reactor, pressurized-water reactor, and coal-fired power plants. The Cascade reactor and blanket are made of ceramic materials and activation is 6 times less than that of the MARS Tandem Mirror Reactor operating at comparable power. Hands-on maintenance of the heat exchangers is possible one day after shutdown. Essentially all tritium is recovered in the vacuum system, with the remainder recovered from the helium power conversion loop. Tritium leakage external to the vacuum system and power conversion loop is only 0.03 Ci/d

  9. 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)

    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

  10. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  11. Oregon State TRIGA reactor power calibration study

    International Nuclear Information System (INIS)

    As a result of a recent review of the Oregon State TRIGA Reactor (OSTR) power calibration procedure, an investigation was performed on the origin and correctness of the OSTR tank factor and the calibration method. It was determined that there was no clear basis for the tank factor which was being used (0.0525 deg. C/kwh) and therefore a new value was calculated (0.0493 deg. C/kwh). The calculational method and likely errors are presented in the paper. In addition, a series of experimental tests were conducted to decide if the power calibration was best performed with or without a mixer, at 100 KW or at 1 MW. The results of these tests along with the final recommendation are presented. (author)

  12. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, K.L.; Ramuhali, P.; Brenchley, D.L.; Coble, J.B.; Hashemian, H.M.; Konnick, R.; Ray, S.

    2012-09-01

    Executive Summary [partial] The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. A workshop was held to gather subject matter experts to develop the NDE R&D Roadmap for Cables. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, and NDE instrumentation development from the U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), universities, commercial NDE service vendors and cable manufacturers, and the Electric Power Research Institute (EPRI).

  13. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  14. Seasonal Flight Activity of the Sugarcane Beetle (Coleoptera: Scarabaeidae) in North Carolina Using Black Light Traps.

    Science.gov (United States)

    Billeisen, T L; Brandenburg, R L

    2016-04-01

    Seasonal flight activity, adult beetle sex count, and egg production were examined in sugarcane beetles Euetheola rugiceps (LeConte) caught in light traps in North Carolina from the fall of 2009 through the summer of 2014. A regression model using variable environmental conditions as predictive parameters was developed to examine the impact of these conditions on flight activity. Depending on flight trap location and sampling years, beetles exhibited an inconsistent flight pattern, with the majority of adults flying in the spring (April-June) and intermittently in the fall (September-October). Our model indicated that larger numbers of adults collected from traps coincided with an increase in average soil temperature. Sugarcane beetles also exhibit a synchronous emergence during both periods of flight activity. Eggs were detected in females collected from light traps every week throughout the entire sampling period. The majority of females produced 7-12 eggs, with most egg production occurring between 15 May and 1 August. The findings of this research provide adult sugarcane beetle emergence and flight behavior information necessary to determine optimal pesticide application timing.

  15. Instrumentation and control for reactor power setback in PFBR

    International Nuclear Information System (INIS)

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

  16. 78 FR 8190 - Commercial Leasing for Wind Power on the Outer Continental Shelf (OCS) Offshore North Carolina...

    Science.gov (United States)

    2013-02-05

    ... Bureau of Ocean Energy Management Commercial Leasing for Wind Power on the Outer Continental Shelf (OCS... Information and Nominations for Commercial Leasing for Wind Power Offshore North Carolina (Call), published on December 13, 2012 (77 FR 7204). DATES: BOEM must receive your nomination describing your interest...

  17. Investigation of the Impact of ENDF/B-VI Cross Sections on the H.B. Robinson-2 Pressure-Vessel Flux Prediction

    Energy Technology Data Exchange (ETDEWEB)

    Remec, I

    1999-06-01

    This report discusses the impact of the change from the SAILOR cross-section library, based on the ENDF/B-IV data, to the BUGLE-96 cross-section library, based on the ENDF/B-VI data, on the neutron flux prediction in the H. B. Robinson-2 pressure vessel, in the surveillance capsule, and in the cavity. The fast flux (E > 1 MeV) from the transport calculations with the BUGLE-96 library is {approximately}6% higher in the surveillance capsule and at the PV inner wall, and {approximately}25% higher in the reactor cavity than the flux from the transport calculations with the SAILOR library. These changes result from the combined effect of the changes in the cross sections, which cause significant increases in the calculated fluxes, and much smaller decreases in the fast fluxes due to the changes in the fission spectra. The increase in the calculated fast flux due to the changes in the cross sections only is {approximately}9% in the capsule and at the pressure vessel (PV) wall, and {approximately}30% in the cavity. The changes in the fission spectra lead to decreases in the order of {approximately}3-4% in calculated fast fluxes.

  18. Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

  19. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Short Description of the Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described, and the radiological consequences of the core modification are quantified to be tolerable

  20. Passive and inherent safety technologies for light-water nuclear reactors

    International Nuclear Information System (INIS)

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs

  1. United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    International Nuclear Information System (INIS)

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

  2. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  3. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available The Consortium for Advanced Simulation of Light Water Reactors (CASL was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major “Challenge Problems” in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  4. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    Science.gov (United States)

    Kulesza, Joel A.; Franceschini, Fausto; Evans, Thomas M.; Gehin, Jess C.

    2016-02-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major "Challenge Problems" in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  5. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  6. Research on physical and chemical parameters of coolant in Light-Water Reactors

    International Nuclear Information System (INIS)

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  7. Reactor/Brayton power systems for nuclear electric spacecraft

    International Nuclear Information System (INIS)

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system

  8. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    Science.gov (United States)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  9. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    International Nuclear Information System (INIS)

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated

  10. Comparison of Pickering NGS performance with world power reactors, 1977

    International Nuclear Information System (INIS)

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

  11. Air quality effects of South Carolina electric and gas company's proposed Cope power plant

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — A review of preliminary determination prepared by the South Carolina bureau of air quality control for the South Carolina electric and gas company's proposed cope...

  12. Testing of a reactimeter for a light water reactor in the range + 500 to - 5000 pcm

    International Nuclear Information System (INIS)

    This apparatus is designed to measure instantaneously the positive or negative reactivity of a uranium reactor moderated by light water, on condition that the point of departure is the critical state of the reactor, or an already known sub-critical state. Slight modifications only are required to adapt it to another type of reactor. It is an analogue computer which simply inverses the transfer function of the reactor; it is not therefore a model reactor of which the output voltage is connected by a servo-mechanism to the power of the reactor to give the reactivity; the principle of the calculation of the reactivity does not depend on a servomechanism. One of its disadvantages is that it cannot operate outside a power variation range of 2.5 decades. However the measurement of a negative reactivity value between 0 and 3000 pcm is immediate. It measures the reactivity without deducting it from the period; it therefore gives the reactivity very precisely both for divergence and convergence even through in this latter case the period does not in fact exist. The equipment makes it possible to calibrate very rapidly the control rods of a reactor (the rod-drop method), to measure the reactivity of an experiment in the core, and to measure certain temperature effects. It is also possible by introducing a control into the core at a measured rate, to deduce directly its efficiency curve. (author)

  13. PR-EDB: Power Reactor Embrittlement Database - Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  14. Assessment of light water reactor accident management programs and experience

    International Nuclear Information System (INIS)

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation

  15. Assessment of light water reactor accident management programs and experience

    Energy Technology Data Exchange (ETDEWEB)

    Hammersley, R.J. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  16. Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 3

    International Nuclear Information System (INIS)

    This report, Supplement No. 3 to the Safety Evaluation Report for the application filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicants) for a license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 and 2

  17. Safety evaluation report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 1

    International Nuclear Information System (INIS)

    This report, Supplement No. 1 to the Safety Evaluation Report for the application filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicant) for license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400), has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

  18. European community light water reactor safety research projects. Experimental issue

    International Nuclear Information System (INIS)

    Research programs on light water reactor safety currently carried out in the European Community are presented. They cover: accident conditions (LOCA, ECCS, core meltdown, external influences, etc...), fault and accident prevention and means of mitigation, normal operation conditions, on and off site implications and equipment under severe accident conditions, and miscellaneous subjects

  19. Risk management and decision rules for light water reactor

    International Nuclear Information System (INIS)

    The process of developing and adopting safety objectives in quantitative terms can provide a basis for focusing societal decision making on the suitability of such objectives and upon questions of compliance with those objectives. A preliminary proposal for a light water reactor (LWR) risk management framework is presented as part of that process

  20. Capital costs of light water reactors: the USA

    International Nuclear Information System (INIS)

    The cost of building a modern nuclear power plant is greater than that of almost any other single civilian project - costs of individual plants are reckoned in hundreds of millions of pounds in the UK, and up to a billion dollars or more in the USA. Hence, depending on the size of nuclear programmes and their funding, escalation of nuclear capital costs may have important economic and social consequences through its effects on overall resource allocation. It is therefore important to analyse the extent and, as far as possible, the sources of cost increases and escalation, in order to see if the experience yields implications for technology policy. The USA has much the greatest experience in nuclear construction: it also has by far the largest amount of published information on the subject of capital costs. As all other countries lack either sufficient experience and/or adequate published cost information, it is impossible to conduct a genuine international comparison, and this paper is confined to an examination of US experience. This paper therefore assembles and evaluates currently available data on light water reactor (PWR and BWR) capital costs in the USA. (author)

  1. LIBRA - a light ion beam fusion conceptual reactor design

    International Nuclear Information System (INIS)

    The LIBRA light ion beam fusion commercial reactor study is a self-consistent conceptual design of a 330 MWe power plant with an accompanying economic analysis. Fusion targets are imploded by 4 MJ shaped pulses of 30 MeV Li ions at a rate of 3 Hz. The target gain is 80, leading to a yield of 320 MJ. The high intensity part of the ion pulse is delivered by 16 diodes through 16 separate z-pinch plasma channels formed in 100 torr of helium with trace amounts of lithium. The blanket is an array of porous flexible silicon carbind tubes with Li17Pb83 flowing downward through them. These tubes (INPORT units) shield the target chamber wall from both neutron damage and the shock overpressure of the target explosion. The target chamber is 'self-pumped' by the target explosion generated overpressure into a surge tank partially filled with Li17Pb83 that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequently without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of a 331 MWe LIBRA design is estimated to be 2843 Dollar/kWe while a 1200 MWe LIBRA design will cost approximately 1300 Dollar/kWe. (orig.)

  2. Impact of inflow transport approximation on light water reactor analysis

    Science.gov (United States)

    Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung

    2015-10-01

    The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.

  3. Multi-Application Small Light Water Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept

  4. Multi-Application Small Light Water Reactor Final Report

    International Nuclear Information System (INIS)

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO2, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as cogeneration

  5. Axial power monitoring uncertainty in the Savannah River Reactors

    International Nuclear Information System (INIS)

    The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

  6. Safety Re-evaluation of Kyoto University Research Reactor by reflecting the Accident of Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, K.; Yamamoto, T. [Kyoto Univ., Kyoto (Japan)

    2013-07-01

    Kyoto University Research Reactor (KUR) is a light-water moderated tank-type reactor operated at rated thermal power of 5MW. After the accident of Fukushima Daiichi nuclear power plant, we have settled a 40-ton water tank near the reactor room, and prepared a mobile fire pump and a mobile power generator as additional safety measures for beyond design basis accidents (BDBAs). We also have conducted the safety re-evaluation of KUR, and confirmed that the integrity of KUR fuels could be kept against the BDBA with the use of the additional safety measures when the several restrictions were imposed on the reactor operation.

  7. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

  8. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    Science.gov (United States)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  9. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  10. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    International Nuclear Information System (INIS)

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant's current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the

  11. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

    2015-06-01

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

  12. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  13. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  14. Waste management in light-water reactors

    International Nuclear Information System (INIS)

    The most important objectives of concentrate and solid waste treatment are reduction of the waste to the smallest volume, radioactive exposure of the personnel of the power plants and outside for operation, handling and transportation, protection against migration of the concentrated radioactive substances after final disposal and observance of shipping requirements, national laws and ministerial waste storage regulations. A variety of technologies is available for the realization of these objectives. Important parameters for the selection and design of concentrate and solid waste treatment processes are waste type, quantity, activity, means for immobilization and the achievable reduction factors. The most important technologies for the treatment of liquid concentrates, combustible and non-combustible solid waste are available for example: In-Drum-Drying, Borate-Solidification (PWR), Drum Drier, Residue Filter Drying, Bituminization, Solidification with cement, Incineration, Shredding, Compacting etc. and of course combinations of the various mentioned procedures which result in the best possible waste disposal for the entire power plant. (orig./RW)

  15. Consumption of the electric power inside silent discharge reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yehia, Ashraf, E-mail: yehia30161@yahoo.com [Department of Physics, Faculty of Science, Assiut University, Assiut 71516, Arab Republic of Egypt and Department of Physics, College of Science and Humanitarian Studies at Alkharj, Salman bin Abdulaziz University, P.O. Box 83, Alkharj 11942 (Saudi Arabia)

    2015-01-15

    An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodes in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.

  16. Economics of advanced light water reactors - Recent update

    International Nuclear Information System (INIS)

    This paper includes recently updated analyses of the economic prospects of advanced light water reactors (ALWRs) during the decade of the 1990s. United Engineers and Constructors (UE and C) has performed engineering economic analyses related to ALWRs over the last 5 yr using both target economics and detailed cost-estimating methodologies. It has been found through such cost comparisons that properly designed and constructed ALWRs should cost less than the target cost figures listed above and significantly less than the pressurized water reactor better experience reference LWR plant cost

  17. Domestic light water reactor fuel design evolution. Volume III

    International Nuclear Information System (INIS)

    Volume III of this report examines the design evolution of domestic light water reactor fuel. The fuel of each vendor is individually described. Tables and figures detail the fuel's design parameters. A data base of this nature is required for the design of an underwater fuel disassembly and rod storage system. An assessment of fuel failure mechanisms and fuel performance is presented showing that spent fuel pool operational problems will be minimal or nonexistent. A summary and projection of spent fuel discharges, organized by reactor and fuel design type, is included to show the magnitude and composition of the spent fuel situation facing the nuclear industry

  18. Power start up of the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

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

  19. Trend of inspection and evaluation, repair and preventive maintenance guidelines for light water reactor internals

    International Nuclear Information System (INIS)

    Inspection and Evaluation Guidelines for Light Water Reactor Internals have been enacted because the damages of stress corrosion cracking etc. were found recently in light water reactor internals. This paper describes the trend of the guidelines. (author)

  20. Burnup analysis of the power reactor, 1

    International Nuclear Information System (INIS)

    Several years of endeavors has been devoted to development of three-dimensional nuclear-thermal-hydro-dynamic simulators and research by basing the progress on the merits and demerits of the variational method, the functional approximation method, etc. As the result, the three-dimensional nuclear-thermal-hydro-dynamic code FLORA has been prepared. It has the following features. (1) The executive time is one third -- half as much as that by the convensional programs. (2) Numerical error is small when neutron spectrum mismatches. (3) In the fuels in which the distributions of Gd2O3 and enrichments are localized axially in the reactor core, three-dimensional nuclear-thermal-hydro-dynamic calculations are possible. (4) The transport kernel can be obtained by the coarse mesh method and the functional approximation method. (5) Albedo can be calculated by the two-group diffusion theory. (6) Power distribution can be obtained in the case of partial control rods inserted in the core. The course taken to the preparation, the theoretical background and example calculations with FLORA are described. The present report can be also used as a manual. (auth.)

  1. Light a CANDLE. An innovative burnup strategy of nuclear reactors

    International Nuclear Information System (INIS)

    CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)

  2. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    The technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and AECL, has been safely,economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world

  3. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  4. Flow-induced vibration for light water reactors. Final progress report, July 1981-September 1981

    International Nuclear Information System (INIS)

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the final period from July 1981 to September 1981. This is the last quarterly progress report to be issued for this program

  5. Utility leadership in reopening the nuclear option with advanced light water reactors

    International Nuclear Information System (INIS)

    Since 1981, the Electric Power Research Institute (EPRI) has been pursing the development of the advanced light water reactor (ALWR). The ALWR Program is comprised of five phases and are described in the paper. In order to meet the anticipated baseline power generation requirements in the US, the Nuclear Power Oversight Committee (NPOC) has developed a strategic plan for ALWR implementation in order to regain the nuclear option in the United States. The paper also covers the policies behind the utility requirements, the status of ALWR developments in the United States, the electricity demands during the period 1990-2010, and some of the innovative features of the passive plants presently under design

  6. 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 ...

  7. The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

    International Nuclear Information System (INIS)

    Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author)

  8. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    Energy Technology Data Exchange (ETDEWEB)

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  9. Nuclear Power Reactors in the World. 2016 Ed

    International Nuclear Information System (INIS)

    Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to reactor units in IAEA Member States. This thirty-sixth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2015. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. Data is collected by the IAEA via designated national correspondents in Member States

  10. POWERED LED LIGHTING SUPPLIED FROM PV CELLS

    Directory of Open Access Journals (Sweden)

    Tirshu M.

    2011-12-01

    Full Text Available The paper deals with practical realization of efficient lighting system based on LED’s of 80W total power mounted on corridor ceiling total length of which is 120m and substitutes existing traditional lighting system consisting of 29 lighting blocks with 4 fluorescent lamps each of them and summary power 2088W. Realized lighting system is supplied from two photovoltaic panels of power 170W. Generated energy by PV cells is accumulated in two accumulators of 75Ah capacity and from battery by means of specialized convertor is applied to lighting system. Additionally, paper present data measured by digital weather station (solar radiation and UV index, which is mounted near of PV cells and comparative analyze of solar energy with real energy generated by PV cells is done. Measured parameters by digital weather station are stored by computer in on-line mode.

  11. LMFBR type reactor and power generation system using the same

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira.

    1994-02-25

    A reactor core void reactivity of a reactor main body is set to negative or zero. A heat insulation structure is disposed on the inner wall surface of a reactor container. Oxide fuels or nitride fuels are used. A fuel pin cladding tube has a double walled structure having an outer side of stainless steel and an inner side of niobium alloy. Upon imaginary event, boiling is allowed. Even if boiling of coolants should occur by temperature elevation of fuels upon imaginary event, since reactor core fuels comprises oxides or nitrides, they have a heat resistance, further, and since the fuel pin cladding tube has super heat resistance, it has a high temperature strength, so that it is not ruptured and durable to the coolant boiling temperature. Since the reactor core void reactivity is negative or zero, the reactor core is in a subcritical state by the boiling, and the reactor core power is reduced to several % of the rated power. Accordingly, boiling and non-boiling are repeated substantially permanently in the reactor core, during which safety can be kept with no operator's handling. Further, heat generated in the reactor core is gradually removed by an air cooling system for the reactor container. (N.H.).

  12. Non-Powered Spectrophotometry for Lighting Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The spectrum of a light affects crew health.  This solution does not require power or a calibration cycle. The premise is that an optical solution is...

  13. Multi-Applications Small Light Water Reactor - NERI Final Report

    Energy Technology Data Exchange (ETDEWEB)

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  14. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  15. Overview of core simulation methodologies for light water reactor analysis

    International Nuclear Information System (INIS)

    The current in-core fuel management calculation methods provide a very efficient route to predict neutronics behavior of light water reactor (LWR) cores and their prediction accuracy for current generation LWRs is generally sufficient. However, since neutronics calculations for LWRs are based on various assumptions and simplifications, we should also recognize many implicit limitations that are 'embedded' in current neutronics calculation methodologies. Continuous effort for improvement of core simulation methodologies is also discussed. (author)

  16. Gas core reactor power plants designed for low proliferation potential

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF/sub 6/ and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on /sup 233/U born from thorium. Fission product removal was continuous. Newly born /sup 233/U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of /sup 233/U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors.

  17. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

  18. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Lingzhi, E-mail: lxia4@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)

    2014-10-15

    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  19. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  20. Safety evaluation report related to the operation of Shearon Harris Nuclear Power Plant, Units 1 and 2. Docket Nos. STN 50-400 and STN 50-401

    International Nuclear Information System (INIS)

    The Safety Evaluation Report for the application filed by the Carolina Power and Light Company, as applicant and owner, for licenses to operate the Shearon Harris Nuclear Power Plant Units 1 and 2 (Docket Nos. 50-400 and 50-401) has been prepared by the Office of Nuclear Reactor Regulation of US Nuclear Regulatory Commission. The facility is located near Raleigh, North Carolina. Subject to favorable resolution of the items discussed in this report, the NRC staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public

  1. Self-operation type power control device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru.

    1993-07-23

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

  2. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  3. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

    International Nuclear Information System (INIS)

    In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

  4. Power instability and stochastic dynamics of periodic pulsed reactors

    International Nuclear Information System (INIS)

    This paper reports that physicists dealing with conventional reactor dynamics recognize two types of instability and reactor behavior beyond the stability region: asymptotic excursions and nonlinear periodic oscillations. A periodically pulsed reactor (PPR) has another peculiar instability: Under certain conditions, its power tends to oscillate at a frequency just twice less than the reactor pulsation frequency. The PPR dynamics far beyond the stability region are analyzed by using a discrete nonlinear model. A PPR with a negative temperature reactivity effect inevitably shows the chaotic power pulse energy behavior known as deterministic chaos. The way by which a reactor goes to chaos is defined by the time dependence of the feedback and by the kind of dynamics model used

  5. Nonthermal plasma reactors for the production of light hydrocarbon olefins from heavy oil

    Directory of Open Access Journals (Sweden)

    G. Prieto

    2003-03-01

    Full Text Available During the last decade, nonthermal plasma technology was applied in many different fields, focusing attention on the destruction of harmful compounds in the air. This paper deals with nonthermal plasma reactors for the conversion of heavy oil into light hydrocarbon olefins, to be employed as gasoline components or to be added in small amounts for the catalytic reduction of nitrogen oxide compounds in the treatment of exhaust gas at power plants. For the process, the plate-plate nonthermal plasma reactor driven by AC high voltage was selected. The reactor was modeled as a function of parameter characteristics, using the methodology provided by the statistical experimental design. The parameters studied were gap distance between electrodes, carrier gas flow and applied power. Results indicate that the reactions occurring in the process of heavy oil conversion have an important selective behavior. The products obtained were C1-C4 hydrocarbons with ethylene as the main compound. Operating the parameters of the reactor within the established operative window of the system and close to the optimum conditions, efficiencies as high as 70 (mul/joule were obtained. These values validate the process as an in-situ method to produce light olefins for the treatment of nitrogen oxides in the exhaust gas from diesel engines.

  6. Technology and use of low power research reactors

    International Nuclear Information System (INIS)

    The report contains a summary of discussions and 10 papers presented at the Consultants' Meeting on the Technology and Use of Low Power Research Reactors organized by the IAEA and held in Beijing (China) during 30 April - 3 May 1985. The following topics have been covered: reactor utilization in medicine and biology, in universities, for training, as a neutron source for radiography and some remarks on the safety of low power research reactors. A separate abstract was prepared for each paper presented at the meeting

  7. Regulation concerning installation and operation of reactors for power generation

    International Nuclear Information System (INIS)

    The regulation is defined under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors and provisions concerning installation and operation of reactors for power generation in the order for execution of the law. Basic concepts and terms are explained, such as: radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; safeguarded area; inspected surrounding area and employee. The application for permission of installation of reactors shall include location and general structure of reactor facilities, structure and equipment of reactors, handling and storing facilities of nuclear fuel materials and facilities for measurement and control, etc. Operation program of reactors shall be prepared for each reactor according to the form attached and filed every fiscal year from that one when the operation is expected to begin. Records shall be made for each reactor and kept for particular periods on inspection of reactor facilities, operation, fuel assembly, control of radiation and maintenance, etc. Entrance to the controlled area shall be limited through specified measures. Exposure dose, inspection, check up, independent examination and operation of reactors, transport and disposal in the works or the enterprise and others are in detail stipulated. Reports shall be submitted to the Minister of International Trade and Industry on concentration of radioactive materials, exposure dose of employees and other designated matters. (Okada, K.)

  8. Analysis of TRIGA reactor thermal power calibration method

    International Nuclear Information System (INIS)

    Analysis of thermal power method of the nuclear instrumentation of the TRIGA reactor in Ljubljana is described. Thermal power calibration was performed at different power levels and at different conditions. Different heat loss processes from the reactor pool to the surrounding are considered. It is shown that the use of proper calorimetric calibration procedure and the use of heat loss corrections improve the accuracy of the measurement. To correct the position of the control rods, perturbation factors are introduced. It is shown that the use of the perturbation factors enables power readings from nuclear instrumentation with accuracy better than without corrections.(author)

  9. Study on a decay heat removal system of light water reactors using air coolers

    International Nuclear Information System (INIS)

    In the present work, a passive decay heat removal system for light water reactors (LWRs) based on a new concept is studied referring to an air cooling system (ACS) of the fast breeder reactor Monju. The present study will contribute to the reduction of severe accident risks of nuclear power plants. In this system, a blower for an air cooler (AC) is operated using the rotation of a small steam turbine by generated steam in order to cool heat transfer tubes by forced convection of air. The purpose of the present work is to investigate the plant transient caused by a station blackout (SBO) using the plant system code NETFLOW++ and decay heat removal characteristics. A calculation model is the Advanced Boiling Water Reactor (ABWR) in Japan. (author)

  10. An overview of future sustainable nuclear power reactors

    Directory of Open Access Journals (Sweden)

    Andreas Poullikkas

    2013-01-01

    Full Text Available In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA. In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will

  11. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  12. Multi-Application Small Light Water Reactor (MASLWR). Annex I

    International Nuclear Information System (INIS)

    The Multi-Application Small Light Water Reactor (MASLWR) design was developed through a collaborative effort sponsored by the nuclear energy research initiative (NERI) programme of the U.S. Department of Energy (DOE). The design and testing team was comprised of staff from the Idaho National Engineering and Environmental Laboratory (INEEL), Oregon State University (OSU), and NEXANT-Bechtel. The primary objectives of the project were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate its technical feasibility by conducting testing in a scaled integral test facility. The design has evolved from an initial concept that employed a primary system layout similar to a traditional pressurized water reactor (PWR), with U-tube steam generators external to the reactor vessel, with the thermal centers of the steam generators elevated to enhance natural circulation. The preliminary estimates for the early design indicated that the busbar cost would be about 5.7 cents/kWh, which is far above the goal of 4 cents/kWh. It was concluded that cost reductions could be achieved by using smaller, simpler, factory-assembled units. Therefore, the focus of the project was redirected to a self-contained modular reactor design consisting of an integrated reactor vessel, steam generator, and high-pressure containment vessel. The entire reactor module would be shop fabricated and transportable to a site on most railways or roads. This sealed modular design eliminates the need for on-site refuelling. The MASLWR concept relies on available nuclear technologies. The nuclear core and turbine generators designs were based on configurations used in a typical PWR. This allows the use of current industry expertise and manufacturing capabilities. The novelty of the system comes from the integration of the entire primary side into a single modular unit with an

  13. Investigation of materials for fusion power reactors

    Science.gov (United States)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  14. Design of megawatt power level heat pipe reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-12

    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

  15. Analysis of Nigeria research reactor-1 thermal power calibration methods

    Energy Technology Data Exchange (ETDEWEB)

    Agbo, Sunday Arome; Ahmed, Yusuf Aminu; Ewa, Ita Okon; Jibrin, Yahaya [Ahmadu Bello University, Zaria (Nigeria)

    2016-06-15

    This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

  16. Small reactor power systems for manned planetary surface bases

    International Nuclear Information System (INIS)

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options

  17. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    International Nuclear Information System (INIS)

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 25800F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

  18. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Richardson, K.D.

    1987-10-01

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.

  19. Sealing device for nuclear power reactor

    International Nuclear Information System (INIS)

    The sealing device is to stop a leak on a reactor pressure vessel where control of the output of reactor is arranged by control rods which are handled by drives connected to control rods and bars in tubes which penetrate the reactor wall. Each tube has a supporting case on the inside of the wall opened to the hole and welded to the tube. The weld may crack and leak. Then an inner sealing tube made of soft metallic material whose outer surface is conical is drawn on to the tube over which an outer sealing tube made of hard metallic material and conical inner surface is placed. On both sides of the crack special adhering planes are formed between the inner sealing tube and the tubes or the supporting case. When the outer sealing tube is pressed over the inner sealing tube, the conical surfaces tighten it against the tube and the supporting case

  20. Plutonium breeding in liquid-metal fast breeder reactors and light water reactors

    International Nuclear Information System (INIS)

    The possibilities of breeding in liquid-metal fast breeder reactors (LMFBRs) and light water reactors (LWRs) are compared in two ways. The feasibility of breeding has been demonstrated in the Phenix reactor with a measured gain of 0.14. The gain in Superphenix will amount to about0.20. The studies show that while maintaining the performance of commercial reactors their breeding gain can be further increased either by the concept of heterogeneous cores or by using carbide or nitride fuel (breeding gain about0.35). Recently, the old idea of breeding in advanced pressurized water reactors (PWRs) has been taken up again with the objective of attaining a gain of 0.05. Unfortunately, these objectives had to be limited to a conversion ratio of 0.9 for safety reasons, and it is not certain whether operation will be rewarding economically. The strategy of substituting PWRs is examined using the French example. By gradually introducing LMFBRs, the cumulated uranium supplies in France can be kept within reasonable limits, which means that they attain three to four times the home resources. This is not possible with advanced LWRs, which can be considered only as a possible backup solution for plutonium recycling into PWRs

  1. Needs of nuclear data for advanced light water reactor

    International Nuclear Information System (INIS)

    Hitachi has been developing medium sized ABWRs as a power source that features flexibility to meet various market needs, such as minimizing capital risks, providing a timely return on capital investments, etc. Basic design concepts of the medium sized ABWRs are 1) using the current ABWR design which has accumulated favorable construction and operation histories as a starting point; 2) utilizing standard BWR fuels which have been fabricated by proven technology; 3) achieving a rationalized design by suitably utilizing key components developed for large sized reactors. Development of the medium sized ABWRs has proceeded in a systematic, stepwise manner. The first step was to design an output scale for the 600MWe class reactor (ABWR-600), and the next step was to develop an uprating concept to extend this output scale to the 900MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. In addition, Hitachi and MHI developed an ultra small reactor, 'Package-Reactor'. About the nuclear data, for the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores. (author)

  2. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  3. Non-linear analysis in Light Water Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  4. Present status of graphite-moderated power reactor decommissioning in foreign countries

    International Nuclear Information System (INIS)

    From 1960's on, graphite-moderated power reactors, being either of CO2 gas cooled or light water cooled type, had opened the nuclear electricity generation worldwide. Such pioneering reactors as UK Magnoxes, French GCRs, Russian AMBs had been operated for more than 20 years up to 40 years. Some of these pioneering power reactors have already been brought into permanent shutdowns, followed by decommissioning activities or preparation of decommissioning projects. On the occasion of the recent start of the decommissioning work at the Tokai Power Station, an overview on progress status in shutdown graphite-moderated power plants in several countries is given. In this report are described strategic aspects and some specific dismantling and waste management methods to be notified in individual decommissioning projects, as in the following. A few UK Magnox power stations have been in preparation for 'Safestore Construction', which will be reserved for more than 100 years after shutdown. The UKAEA's WAGR has been long undertaken as one of the big EC's reactor decommissioning projects, with extensive R and D work carried out for immediate dismantling of the graphite-moderated reactor. The recent successful progresses have revealed safe and commercial-scale dismantling procedures and technologies, which may facilitate an early dismantling shutdown nuclear facilities. The French GCR plants have been in plant-by-plant preparation for safestore for 30-40 years. The Spanish Vandellos-1 and Italian Latina plants are also under decommissioning operations similarly as in UK and France. All experimental and prototype high temperature reactor plants in Germany and USA had already been under decommissioning processes, with various safestore conditions depending on the specific project circumstances. The German AVR is being prepared for step-by-step dismantling the reactor structure. The Beloyarsk NPP based on ex-Soviet Union graphite reactor concept is still in preparatory phase in

  5. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  6. Evolutionary/advanced light water reactor data report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  7. Materials Inventory Database for the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  8. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

  9. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  10. Laser fusion power reactor system (LFPRS)

    Energy Technology Data Exchange (ETDEWEB)

    Kovacik, W. P.

    1977-12-19

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements. (MOW)

  11. Special Nuclear Material Control by the Power Reactor Operator

    International Nuclear Information System (INIS)

    A relatively new and extremely valuable fuel for electric power production, uranium, requires very careful inventory control from the time the reactor operator assumes financial responsibility for this material until, as partially expended fuel, it is transferred to another facility and the remaining part of its initial value is recovered. Most power reactor operators were operating fossil-fuelled power plants before the advent of nuclear power and have long since established rather complete and adequate controls for these fossil fuels. The reactor operator must have no less adequate controls for the special nuclear material used in his nuclear plant. Power reactor, operation is not an ancient science and during its relatively short history our engineers and scientists have been constantly improving plant designs and methods of operation to reduce costs and make our nuclear plants competitive with fossil-fuelled conventional plants. Nuclear material management must be as modern and efficient as is humanly possible to ensure that technological advances leading to reduced costs are not lost by poor handling of nuclear fuel and the records pertaining to fuel inventory. Nuclear material management requires the maintaining of complete and informative records by the power reactor operator. These records need not be complex to satisfy the criteria of completeness and adequacy. In fact, simplicity is extremely desirable. Despite the fact that nuclear fuel is new and completely different to our conventional fuels no mystery should be attached thereto. Nuclear material control as part of nuclear material management is not limited to simple inventory work but it is the basis for a great deal of other activity that is an inherent part of any power reactor operations such as irradiated fuel shipments, reprocessing of spent fuel, with its associated accounting for reclaimed fuel and material produced during reactor operation, and the establishing and maintaining of an adequate

  12. Lighting: The Killer App of Village Power

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This paper looks at lighting systems as the major market for village level power generation. To the consumer it is something which is needed, could come from a much friendlier source, and the issues of affordability, convenience, and reliability are important. To the supplier lighting has an enormous range of potential customers, it opens the opportunity for other services, and even small demand can give big returns. Because the efficiency of the light source is critical to the number of lights which a fixed power supply can drive, it is important to pick the proper type of bulb to use in this system. The paper discusses test results from an array of fluorescent and incadescent lamps, compared with a kerosene lamp. Low wattage fluorescents seem to perform the best.

  13. Power Reactor Fuel Reprocessing: Mechanical Phase

    Energy Technology Data Exchange (ETDEWEB)

    Klima, B. B.

    1959-07-01

    The major events in the mechanical phase of the Power Reactor fuels reprocessing program during June were: 1. Feasibility of shearing of fuel elements without disassembly has been demonstrated in tests using porcelain-loaded prototype fuel elements. 2. Further work with the Manco shear was not deemed tb be advisable since permission has been granted to use another shear for cutting UO{sub 2}-loaded fuel elements. 3. Necessity to strip the windows in Building 3048, to sandblast, and repaint them has seriously disrupted occupancy of the cell by July 1. Start of installation probably will not be before August 1. 4. A cold SRE element should be received during July which will permit a direct look a t the problems associated with processing of these irradiated fuel elements. 5. Concurrence with AEC, Atomics International, and ORNL people on the fabrication of a poisoned carrier was obtained and all criteria for the carrier were released and the design was completed. 6. A decision was made to install and use a 24-inch Ty-Sa-Man saw which is on hand and was originally purchased for use in the Segmenting Facility for the SRE reprocessing. This will be used instead of the multipurpose saw to allow more time to refine the design of that saw. The multipurpose saw will be installed for use in subsequent reprocessing programs. This report will chronicle the changes in status which occurred during the calendar month of June. A complete description of each item is not included and may be found in the parent report. The dates indicated on the schedule have slipped since the last report primarily due to increase in scope of the work and postponement on all phases of the work except for the SRE preparations. Twenty-four new items have been added to the schedule. The status of procurement is shown. A total of 93 purchase requests have been turned in to t% Purchasing Department. A total of $199,261.83 has been committed by purchase orders, and a total of 56 purchase orders have been

  14. Research Reactor Power Control System Design by MATLAB/SIMULINK

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yong Suk; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Im, Ki Hong [Samsung Electronics, Suwon (Korea, Republic of)

    2013-07-01

    In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure.

  15. Research Reactor Power Control System Design by MATLAB/SIMULINK

    International Nuclear Information System (INIS)

    In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure

  16. Total loss of AC power analysis for EPR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Darnowski, Piotr, E-mail: piotr.darnowski@itc.pw.edu.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Skrzypek, Eleonora, E-mail: eleonora.skrzypek@ncbj.gov.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); National Centre for Nuclear Research (NCBJ), A. Sołtana 7, 05-400 Otwock-Świerk (Poland); Mazgaj, Piotr, E-mail: piotr.mazgaj@itc.pw.edu.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Świrski, Konrad [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Gandrille, Pascal [AREVA NP SAS, Tour AREVA, 1 place Jean Millier, 92084 Paris La Défense (France)

    2015-08-15

    Highlights: • Total loss of AC power (Station Blackout) was simulated for the EPR reactor model. • In-vessel phase of the accident is under consideration. • Comparison of MELCOR and MAAP results is presented. • MELCOR and MAAP results are comparable. - Abstract: In this paper the results of severe accident simulations for the EPR reactor in the case of loss of offsite power combined with total failure of all diesel generators (total loss of AC power) are presented. Calculations were performed with MELCOR 2.1 computer code for in-vessel phase of the accident. In this scenario, the unavailability of all offsite and onsite power sources and the lack of cooling leads directly to core degradation, material relocation to the lower plenum and rupture of the reactor pressure vessel. MELCOR results were compared qualitatively and quantitatively with MAAP4 code results and show a good agreement.

  17. Protective actions as a factor in power reactor siting

    Energy Technology Data Exchange (ETDEWEB)

    Gant, K.S.; Schweitzer, M.

    1984-06-01

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables.

  18. Protective actions as a factor in power reactor siting

    International Nuclear Information System (INIS)

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables

  19. Spent fuel data base: commercial light water reactors

    International Nuclear Information System (INIS)

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel

  20. Environment sensitive cracking in light water reactor pressure boundary materials

    International Nuclear Information System (INIS)

    The purpose of the paper is to review the available methods and the most promising future possibilities of preventive maintenance to counteract the various forms of environment sensitive cracking of pressure boundary materials in light water reactors. Environment sensitive cracking is considered from the metallurgical, mechanical and environmental point of view. The main emphasis is on intergranular stress corrosion cracking of austenitic stainless steels and high strenght Ni-base alloys, as well as on corrosion fatigue of low alloy and stainless steels. Finally, some general ideas how to predict, reduce or eliminate environment sensitive cracking in service are presented

  1. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  2. Nuclear safety. Concerns about the nuclear power reactors in Cuba

    International Nuclear Information System (INIS)

    In 1976, the Soviet Union and Cuba concluded an agreement to construct two 440-megawatt nuclear power reactors near Cienfuegos on the south central coast of Cuba, about 180 miles south of Key West, Florida. The construction of these reactors, which began around 1983, was a high priority for Cuba because of its heavy dependence on imported oil. Cuba is estimated to need an electrical generation capacity of 3,000 megawatts by the end of the decade. When completed, the first reactor unit would provide a significant percentage (estimated at over 15 percent) of Cuba's need for electricity. It is uncertain when Cuba's nuclear power reactors will become operational. On September 5, 1992, Fidel Castro announced the suspension of construction at both of Cuba's reactors because Cuba could not meet the financial terms set by the Russian government to complete the reactors. Cuban officials had initially planned to start up the first of the two nuclear reactors by the end of 1993. However, before the September 5 announcement, it was estimated that this reactor would not be operational until late 1995 or early 1996. The civil construction (such as floors and walls) of the first reactor is currently estimated to be about 90 percent to 97 percent complete, but only about 37 percent of the reactor equipment (such as pipes, pumps, and motors) has been installed. The civil construction of the second reactor is about 20 percent to 30 percent complete. No information was available about the status of equipment for the second reactor. According to former Cuban nuclear power and electrical engineers and a technician, all of whom worked at the reactor site and have recently emigrated from Cuba, Cuba's nuclear power program suffers from poor construction practices and inadequate training for future reactor operators. One former official has alleged, for example, that the first reactor containment structure, which is designed to prevent the accidental release of radioactive material into

  3. Fast-power-reactor optimization by the game theory

    International Nuclear Information System (INIS)

    In the first stage of the use of fast breeder reactor - because fissile-material amounts are small - we are interested in fast breeder reactors which achieve minimum fissile-material mass, with maximum power. This problem shows a two-matrix-game structure. First, we determine a competive-game solution and second, a cooperative-game solution, obtaining in this way the optimum distribution of the fissile and fertile materials in the multizone fast reactors. Another optimization problem which is solved in this paper is finding the reactor structure for which the power non-uniformity factor and the flux non-uniformity factor are minimum. This is, also, a mathematical two-matrix game and it is solved as above. The two optimization problems have different solutions. (author)

  4. The optimum shielding for a power reactor using local components

    International Nuclear Information System (INIS)

    Some local concrete mixtures have been picked out (selected) to be studied as shielding concrete for prospective nuclear power reactor in Syria. This research has interested in the attenuation of gamma radiation and neutron fluxes by these local concretes in the ordinary conditions. In addition to the heat effect on the shielding and physical properties of local concrete. Furthermore the neutron activation of the elements of the local concrete mixtures have been studied that for selection the low-activation materials (low dose rate and short half life radioisotopes). In this way biological shielding for nuclear reactor can be safe during operation of nuclear power reactor, in addition to be low radioactive waste after decommissioning the reactor. (author)

  5. Safeguarding nuclear power stations

    International Nuclear Information System (INIS)

    The basic features of nuclear fuel accounting and control in present-day power reactors are considered. Emphasis is placed on reactor operations and spent-fuel characteristics for Light-Water Reactors (LWRs) and Heavy-Water Reactors (HWRs)

  6. Wind Powering America: A New Wind Economy for South Carolina and Georgia Final Report

    Energy Technology Data Exchange (ETDEWEB)

    SC Energy Office: Southern Alliance for Clean Energy

    2013-02-12

    This report describes all activities undertaken by the Southern Alliance for Clean Energy (SACE) in cooperation with the states of Georgia and South Carolina to develop a public outreach program, including shared analytical and reference tools and other technical assistance.

  7. Heavy-Water Power Reactors. Proceedings Of A Symposium

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  8. Wirelessly powered ultraviolet light emitting diodes for photocatalytic oxidation

    NARCIS (Netherlands)

    Kuipers, J.; Bruning, H.; Yntema, D.R.; Rijnaarts, H.H.M.

    2015-01-01

    A method is presented to distribute small scale light sources in a photocatalytic slurry reactor. The goal of distributing the light sources is to increase photon transfer efficiency, and thereby increasing the reaction rate, compared to using one single light source. The light sources used in this

  9. TREAT light water reactor source term experiments program

    International Nuclear Information System (INIS)

    Four experiments are being conducted in the TREAT facility to investigate the behavior of fission products released from typical LWR fuel overheated to the point of catastrophic cladding degradation. Heatup and steam flow transients are used that simulate the conditions expected in operating power reactors undergoing various types of hypothetical severe accidents. The experiments are integral in nature and are aimed at the physicochemical characterization, near the point of origin, of the biologically important volatile fission products released early in such accidents. Detailed program objectives are discussed, a test matrix is presented, and the test apparatus is described. Pretest analysis and preliminary results are reported for the first test

  10. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    Science.gov (United States)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors. Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat. The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  11. The U.S.-Russian joint studies on using power reactors to disposition surplus weapons plutonium as spent fuel

    International Nuclear Information System (INIS)

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years

  12. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chebeskov, A.; Kalashnikov, A. [State Scientific Center, Obninsk (Russian Federation). Inst. of Physics and Power Engineering; Bevard, B.; Moses, D. [Oak Ridge National Lab., TN (United States); Pavlovichev, A. [State Scientific Center, Moscow (Russian Federation). Kurchatov Inst.

    1997-09-01

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  13. Experimental studies on catalytic hydrogen recombiners for light water reactors

    International Nuclear Information System (INIS)

    In the course of core melt accidents in nuclear power plants a large amount of hydrogen can be produced and form an explosive or even detonative gas mixture with aerial oxygen in the reactor building. In the containment atmosphere of pressurized water reactors hydrogen combines a phlogistically with the oxygen present to form water vapor even at room temperature. In the past, experimental work conducted at various facilities has contributed little or nothing to an understanding of the operating principles of catalytic recombiners. Hence, the purpose of the present study was to conduct detailed investigations on a section of a recombiner essentially in order to deepen the understanding of reaction kinetics and heat transport processes. The results of the experiments presented in this dissertation form a large data base of measurements which provides an insight into the processes taking place in recombiners. The reaction-kinetic interpretation of the measured data confirms and deepens the diffusion theory - proposed in an earlier study. Thus it is now possible to validate detailed numeric models representing the processes in recombiners. Consequently the present study serves to broaden and corroborate competence in this significant area of reactor technology. In addition, the empirical knowledge thus gained may be used for a critical reassessment of previous numeric model calculations. (orig.)

  14. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  15. Transportation and storage of foreign spent power reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-30

    This report describes the generic actions to be taken by the Department of Energy, in cooperation with other US government agencies, foreign governments, and international organizations, in support of the implementation of Administration policies with respect to the following international spent fuel management activities: bilateral cooperation related to expansion of foreign national storage capacities; multilateral and international cooperation related to development of multinational and international spent fuel storage regimes; fee-based transfer of foreign spent power reactor fuel to the US for storage; and emergency transfer of foreign spent power reactor fuel to the US for storage.

  16. Transportation and storage of foreign spent power reactor fuel

    International Nuclear Information System (INIS)

    This report describes the generic actions to be taken by the Department of Energy, in cooperation with other US government agencies, foreign governments, and international organizations, in support of the implementation of Administration policies with respect to the following international spent fuel management activities: bilateral cooperation related to expansion of foreign national storage capacities; multilateral and international cooperation related to development of multinational and international spent fuel storage regimes; fee-based transfer of foreign spent power reactor fuel to the US for storage; and emergency transfer of foreign spent power reactor fuel to the US for storage

  17. Multivariable robust control of an integrated nuclear power reactor

    OpenAIRE

    Etchepareborda A.; Flury C.A.

    2002-01-01

    The design of the main control system of the CAREM nuclear power plant is presented. This plant is an inherently safe low-power nuclear reactor with natural convection on the primary coolant circuit and is self-pressurized with a steam dome on the top of the pressure vessel (PV). It is an integrated reactor as the whole primary coolant circuit is within the PV. The primary circuit transports the heat to the secondary circuit through once-through steam generators (SG). There is a feedwater val...

  18. Detecting, locating and identifying failed fuel in Canadian power reactors

    International Nuclear Information System (INIS)

    This document summarizes how defected fuel elements are detected, located and identified in Canadian CANDU power reactors. Fuel defects are detected by monitoring the primary coolant for gaseous fission products and radioiodines, while location in core is usually performed on-power by delayed neutron monitoring of coolant samples from individual fuel channels or off-power by gamma-ray monitoring of the channel feeder pipes. The systems and techniques used to detect and locate defected fuel in both Ontario Hydro and CANDU 6 power stations are described, along with examples provided by station experience. The ability to detect and locate defected fuel in power stations was greatly enhanced by a fundamental R and D program, which provided an understanding and models of fission-product release and transport, and the post-defect deterioration of failed fuel. Techniques and equipment used to identify and store defected fuel after it has been discharged from the reactor are briefly reviewed

  19. Source driven breeding thermal power reactors, Pt. 2. Using lithium-free neutron sources

    International Nuclear Information System (INIS)

    The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that of the DT device. The efficiency of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed

  20. Typical technology of mechanics on Gen-III passive NPPs and Gen-IV advanced supercritical light water reactors

    International Nuclear Information System (INIS)

    Full text: Technical requirements for Gen-III advanced nuclear power plants, which take passive reactors as the main body, were originally brought forward in American 'Advanced Light Water Reactor Utility Requirement Document' (ALWR-URD) in early 1990's. The primary characteristic of passive nuclear power plant is large amount of simplification to the original active safety systems, replacing or supplementing them with passive safety systems, which enhances safety and economy. However, the replacement of active safety systems by passive safety systems also brings about some mechanics that compel attention, typically, such as load-carrying capability evaluation for steel containment, in-vessel retention (IVR) of molten core debris, seismic design without OBE, thermo-hydraulic issues concerning with coupling between two-phase fluid and solid, etc. At the beginning of this century, six typical Gen-IV advanced reactor types (Sodium Cooled Fast Reactor, Supercritical Water-Cooled Reactor, etc.) were put forward. Among these types of reactors, Supercritical Water-Cooled Reactor adopts supercritical water as coolant and operates above the thermodynamic critical point of water by increasing temperature and pressure of the coolant, which makes the plant economic and efficient. However, this type of reactor also brings about some mechanical difficulties (e.g. pressure fluctuation caused by the supercritical fluid in the core, creep of materials working at high temperature, etc.) for the design of facility and components. In this paper, the issues mentioned above are outlined for further consideration. (author)

  1. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

  2. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  3. The Jefferson Lab High Power Light Source

    Energy Technology Data Exchange (ETDEWEB)

    James R. Boyce

    2006-01-01

    Jefferson Lab has designed, built and operated two high average power free-electron lasers (FEL) using superconducting RF (SRF) technology and energy recovery techniques. Between 1999-2001 Jefferson Lab operated the IR Demo FEL. This device produced over 2 kW in the mid-infrared, in addition to producing world record average powers in the visible (50 W), ultraviolet (10 W) and terahertz range (50 W) for tunable, short-pulse (< ps) light. This FEL was the first high power demonstration of an accelerator configuration that is being exploited for a number of new accelerator-driven light source facilities that are currently under design or construction. The driver accelerator for the IR Demo FEL uses an Energy Recovered Linac (ERL) configuration that improves the energy efficiency and lowers both the capital and operating cost of such devices by recovering most of the power in the spent electron beam after optical power is extracted from the beam. The IR Demo FEL was de-commissioned in late 2001 for an upgraded FEL for extending the IR power to over 10 kW and the ultraviolet power to over 1 kW. The FEL Upgrade achieved 10 kW of average power in the mid-IR (6 microns) in July of 2004, and its IR operation currently is being extended down to 1 micron. In addition, we have demonstrated the capability of on/off cycling and recovering over a megawatt of electron beam power without diminishing machine performance. A complementary UV FEL will come on-line within the next year. This paper presents a summary of the FEL characteristics, user community accomplishments with the IR Demo, and planned user experiments.

  4. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor

    International Nuclear Information System (INIS)

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC's ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC's preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant's research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified

  5. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO2 in stainless steel, of UO2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  6. Background Radiation Measurements at High Power Research Reactors

    CERN Document Server

    Ashenfelter, J; Baldenegro, C X; Band, H R; Barclay, G; Bass, C D; Berish, D; Bowden, N S; Bryan, C D; Cherwinka, J J; Chu, R; Classen, T; Davee, D; Dean, D; Deichert, G; Dolinski, M J; Dolph, J; Dwyer, D A; Fan, S; Gaison, J K; Galindo-Uribarri, A; Gilje, K; Glenn, A; Green, M; Han, K; Hans, S; Heeger, K M; Heffron, B; Jaffe, D E; Kettell, S; Langford, T J; Littlejohn, B R; Martinez, D; McKeown, R D; Morrell, S; Mueller, P E; Mumm, H P; Napolitano, J; Norcini, D; Pushin, D; Romero, E; Rosero, R; Saldana, L; Seilhan, B S; Sharma, R; Stemen, N T; Surukuchi, P T; Thompson, S J; Varner, R L; Wang, W; Watson, S M; White, B; White, C; Wilhelmi, J; Williams, C; Wise, T; Yao, H; Yeh, M; Yen, Y -R; Zhang, C; Zhang, X

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including $\\gamma$-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  7. Computational design of parameters of IRT-2M fuel with enrichment below 20% for low power research reactors

    International Nuclear Information System (INIS)

    This article focuses briefly on characteristics of a possible procedure during reduction of fuel enrichment of two research reactors in the Czech Republic, i.e., LVR-15 research reactor (power up to 15 MW) at NRI Rez and VR-1 training reactor (power up to 5 kW) at CTU Prague. Both reactors are now operating with fuel enriched to 36% of 235U. While the LVR-15 reactor uses Russian IRT-2M fuel, the VR-1 reactor has been operating on IRT-3M fuel for five years already. The goal for both reactors until now was to use Russian IRT-4M fuel with 235U enrichment below 20%. The original idea that the LVR-15 reactor would go through the IRT-3M fuel during the transition to IRT-4M fuel now seems baseless. The article hence shows a possible solution to the current situation for the VR-1 reactor. A convenient solution (based on consultations with the Russian producer) could be a preparation of fuel of IRT-2M geometry with enrichment to 20% of 235U. Such a fuel would not be intended for power research reactors in the first step but for reactors with power up to 100 - 200 kW. The article presents a proposal of this fuel (said proposal was created on the basis of many years' standing experience of operation at VR-1 reactor) and verifying calculations for selected configurations. As for enrichment, matrix, and content of uranium the proposal is based on verified capability of the Russian producer. Emphasis is placed on the necessity of the fuel having a long lifetime in the light water reactors. (author)

  8. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  9. Power flow control using distributed saturable reactors

    Science.gov (United States)

    Dimitrovski, Aleksandar D.

    2016-02-13

    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  10. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Safety Report Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described together with its assemblies and their loading procedure. The content of radioactive materials and the irradiation protection measures are discussed and those accidents are describe in an enveloping manner, from which an influence of the core modification cannot be excluded. Finally, both core versions (Mark-I and Mark-Ia) are compared with each other

  11. NSR-77: a computer code for transient analysis of a light water reactor fuel rod

    International Nuclear Information System (INIS)

    This report describes computer code NSR-77 written in FORTRAN IV for FACOM-M 200 computer in detail. It has been developed for transient response analysis of a light water reactor fuel rod during an accident such as a reactivityy initiated accident, a loss-of-coolant accident or a power-cooling-mismatch accident. The code consists of subcodes which calculate heat conduction in a fuel rod, gas gap conductance between fuel and cladding, heat transfer from cladding to coolant, fluid hydrodynamics, elastic-plastic fuel and cladding deformation, and material properties, and so on. (author)

  12. Electric power from near-term fusion reactors

    International Nuclear Information System (INIS)

    Near-term fusion reactors such as FED or INTOR will probably have primary cooling systems which operate at temperatures lower than is optimal for power production using a conventional steam cycle. This limitation may be imposed by uncertainties in materials behavior or structural limitations. There are economic motivations to demonstrate electric power production from fusion at the earliest possible date. A greater motivation is the elicitation of public interest in and support of fusion as a viable power source. This paper examines requirements and possibilities 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 thermodynamic 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-term reactors. It is possible to use the FED shield in a binary power cycle, and results are presented of thermodynamic analyses of this system. Thermal storage is accomplished by using the latent heat of fusion of a PbBi eutectic. The secondary loop can use R-11, R-113, or hexafluorobenzene as a working fluid. Such a system would cost about $50 million and would generate about 10 MW of electric power

  13. Hydrogen considerations in light-water power reactons

    International Nuclear Information System (INIS)

    A critical review of the literature now available on hydrogen considerations in light-water power reactors (LWRs) and a bibliography of that literature are presented. The subject matter includes mechanisms for the generation of hydrogen-oxygen mixtures, a description of the fundamental properties of such mixtures, and their spontaneous ignition in both static and dynamic systems. The limits for hydrogen flammability and flame propagation are examined in terms of the effects of pressure, temperature, and additives; the emphasis is on the effects of steam and water vapor. The containment systems for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) are compared, and methods to control hydrogen and oxygen under the conditions of both normal operation and postulated accidents are reviewed. It is concluded that hydrogen can be controlled so that serious complications from the production of hydrogen will not occur. The bibliography contains abstracts from the computerized files of the Nuclear Safety Information Center. Key-word, author, and permuted-title indexes are provided. The bibliography includes responses to questions asked by the U. S. Nuclear Regulatory Commission (NRC) which relate to hydrogen, as well as information on normal operations and postulated accidents including generation of hydrogen from core sprays. Other topics included in the ten sections of the bibliography are metal-water reactions, containment atmosphere, radiolytic gas, and recombiners

  14. Design considerations for an inertial confinement fusion reactor power plant

    International Nuclear Information System (INIS)

    To further define the engineering and economic concerns for inertial confinement fusion reactors (ICR's), a conceptual design study was performed by Bechtel Group Incorporated under the direction of Lawrence Livermore National Laboratory (LLNL). The study examined alternatives to the LLNL HYLIFE concept and expanded the previous balance of plant design to incorporate information from recent liquid metal cooled fast breeder reactor (LMFBR) power plant studies. The majority of the effort was to incorporate present laser and target physics models into a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the LLNL JADE concept. In addition to producing a power plant design for LLNL using the JADE example, Bechtel has also examined the applicability of the EAGLE (Energy Absorbing Gas Lithium Ejector) concept

  15. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs

  16. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  17. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

    2009-06-15

    This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

  18. Reliability assurance programme guidebook for advanced light water reactors

    International Nuclear Information System (INIS)

    To facilitate the implementation of reliability assurance programmes (RAP) within future advanced reactor programmes and to ensure that the next generation of commercial nuclear reactors achieves the very high levels of safety, reliability and economy which are expected of them, in 1996, the International Atomic Energy Agency (IAEA) established a task to develop a guidebook for reliability assurance programmes. The draft RAP guidebook was prepared by an expert consultant and was reviewed/modified at an Advisory Group meeting (7-10 April 1997) and at a consults meeting (7-10 October 1997). The programme for the RAP guidebook was reported to and guided by the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR). This guidebook will demonstrate how the designers and operators of future commercial nuclear plants can exploit the risk, reliability and availability engineering methods and techniques developed over the past two decades to augment existing design and operational nuclear plant decision-making capabilities. This guidebook is intended to provide the necessary understanding, insights and examples of RAP management systems and processes from which a future user can derive his own plant specific reliability assurance programmes. The RAP guidebook is intended to augment, not replace, specific reliability assurance requirements defined by the utility requirements documents and by individual nuclear steam supply system (NSSS) designers. This guidebook draws from utility experience gained during implementation of reliability and availability improvement and risk based management programmes to provide both written and diagrammatic 'how to' guidance which can be followed to assure conformance with the specific requirements outlined by utility requirements documents and in the development of a practical and effective plant specific RAP in any IAEA Member State

  19. Effective use of uranium resources in light water reactor system

    International Nuclear Information System (INIS)

    We have proposed an idea of recycling uranium recovered from spent fuels of light water reactors (LWRs), where the recovered uranium is to be re-enriched by a centrifuge cascade conventionally treating natural uranium. The idea is of making it possible to reuse the fuels reproduced in a multi-cycle of re-enrichment. The uranium recycle not only economizes on uranium resources but also gets rid of accumulation of spent fuel masses. In this work, we consider additional processes for effective use of uranium, which are of re-enriching the depleted uranium. The still-more-depleted uranium is advantageous as the matrix of MOX fuels used in LWRs for the purpose of surplus plutonium disposition, because a decrease of 235U in MOX fuel is made up by increasing a dose of plutonium. However, the depleted uranium derived from the cascade enriching the recovered uranium issues a little troublesome problem of 236U concerning its existence and deliveries to the product and the waste. We made a work to investigate the burn-up performance of these remade uranium fuels in model reactors of 1.1GWe-grade PWR and the mass balance in fuel recycles. The results suggest a strategy of effective use of uranium resources in the LWR system. (author)

  20. Supernova Light Curves Powered by Fallback Accretion

    OpenAIRE

    Dexter, Jason; Kasen, Daniel

    2012-01-01

    Some fraction of the material ejected in a core collapse supernova explosion may remain bound to the compact remnant, and eventually turn around and fall back. We show that the late time (> days) power associated with the accretion of this "fallback" material may significantly affect the optical light curve, in some cases producing super-luminous or otherwise peculiar supernovae. We use spherically symmetric hydrodynamical models to estimate the accretion rate at late times for a range of pro...

  1. The JLab high power ERL light source

    Energy Technology Data Exchange (ETDEWEB)

    G.R. Neil; C. Behre; S.V. Benson; M. Bevins; G. Biallas; J. Boyce; J. Coleman; L.A. Dillon-Townes; D. Douglas; H.F. Dylla; R. Evans; A. Grippo; D. Gruber; J. Gubeli; D. Hardy; C. Hernandez-Garcia; K. Jordan; M.J. Kelley; L. Merminga; J. Mammosser; W. Moore; N. Nishimori; E. Pozdeyev; J. Preble; R. Rimmer; Michelle D. Shinn; T. Siggins; C. Tennant; R. Walker; G.P. Williams and S. Zhang

    2005-03-19

    A new THz/IR/UV photon source at Jefferson Lab is the first of a new generation of light sources based on an Energy-Recovered, (superconducting) Linac (ERL). The machine has a 160 MeV electron beam and an average current of 10 mA in 75 MHz repetition rate hundred femtosecond bunches. These electron bunches pass through a magnetic chicane and therefore emit synchrotron radiation. For wavelengths longer than the electron bunch the electrons radiate coherently a broadband THz {approx} half cycle pulse whose average brightness is > 5 orders of magnitude higher than synchrotron IR sources. Previous measurements showed 20 W of average power extracted[1]. The new facility offers simultaneous synchrotron light from the visible through the FIR along with broadband THz production of 100 fs pulses with >200 W of average power. The FELs also provide record-breaking laser power [2]: up to 10 kW of average power in the IR from 1 to 14 microns in 400 fs pulses at up to 74.85 MHz repetition rates and soon will produce similar pulses of 300-1000 nm light at up to 3 kW of average power from the UV FEL. These ultrashort pulses are ideal for maximizing the interaction with material surfaces. The optical beams are Gaussian with nearly perfect beam quality. See www.jlab.org/FEL for details of the operating characteristics; a wide variety of pulse train configurations are feasible from 10 microseconds long at high repetition rates to continuous operation. The THz and IR system has been commissioned. The UV system is to follow in 2005. The light is transported to user laboratories for basic and applied research. Additional lasers synchronized to the FEL are also available. Past activities have included production of carbon nanotubes, studies of vibrational relaxation of interstitial hydrogen in silicon, pulsed laser deposition and ablation, nitriding of metals, and energy flow in proteins. This paper will present the status of the system and discuss some of the discoveries we have made

  2. Power stabilization in CREN-K TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    In order to eliminate power oscillations in the TRIGA MARK II reactor at the 'Centre Regional d'Etudes Nucleaires de Kinshasa' (CREN-K), Zaire, specially made adapters were put around the control rods in the top grid plate. The paper briefly describes how investigations were made to find out the basic reason of the power oscillations and the way these adapters were conceived and installed. (author)

  3. Fast neutron reactor fuel elements and power grid duty cycling

    International Nuclear Information System (INIS)

    The PHENIX power grid cycling operation in 1982-1983 will allow verification of the models and criteria developed in the interim. It will provide indispensible statistical data and will open the way to power grid duty for Super PHENIX beginning in 1986. Although at the present time it is impossible to resolve the question of weekly or daily load variations, it is felt that fast neutron reactor fuel subassemblies should provide satisfactory performance for primary and secondary frequency adjustments

  4. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    International Nuclear Information System (INIS)

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

  5. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  6. Improvement of maintenance at French power reactors

    International Nuclear Information System (INIS)

    By the end of 1984 an availability of 83.2% had been achieved with the thirty-one 900 MW reactors operated by the Electricite de France (EDF) since the beginning of commercial operation. Outages occurred for the following reasons: shutdowns for refuelling and other planned shutdowns, repairs following incidents, and trips. Improvements in maintenance have an impact mainly on the first two causes mentioned above, and their principal objectives are to reduce plant outages, to cut down the overall dosimetry costs involved in interventions, and to improve the quality of the interventions. This is achieved by staff training and feedback from plant outages. Among the resources employed, mention should be made in particular of the methods used for preparing shutdowns and monitoring shutdown plans, which have reached an advanced stage of development. EDF has developed or arranged for the development of much special equipment for routine maintenance, control operations and repairs on PWR plant components; a management catalogue lists 132 operational pieces of equipment and 20 under development. (author)

  7. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  8. Study of the power margins of RBMK-1000 reactor when operating at nominal power

    International Nuclear Information System (INIS)

    This work presents a reliability and control study of the RBMK-1000 reactor. This study consists in determining the power margin of this plant when operating at nominal power in comparison with the physical phenomena limiting power extraction by a coolant fluid.49 figs., 8 tabs., 11 refs. (author)

  9. Estimates of power requirements for a manned Mars rover powered by a nuclear reactor

    Science.gov (United States)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    1991-01-01

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are met using an SP-100 type reactor. The primary electric power needs, which include 30-kWe net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine (FPSE) yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle (CBC) using He/Xe as the working fluid. The specific mass of the nuclear reactor power systrem, including a man-rated radiation shield, ranged from 150-kg/kWe to 190-kg/kWe and the total mass of the Rover vehicle varied depend upon the cruising speed.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. Characterization of radioactive waste from nuclear power reactors

    International Nuclear Information System (INIS)

    Different kinds of radioactive waste are generated as result of the operation of nuclear power reactors and in all cases the activity concentration of several radionuclides had to be determined in order to optimize resources, particularly when dealing with final disposal or long-term storage. This paper describes the three basic approaches usually employed for characterizing nuclear power reactor wastes, namely the direct methods, the semi-empirical methods and the analytical methods. For some radionuclides or kind of waste, the more suitable method or combination of methods applicable is indicated, stressing that these methods shall be developed and applied during the waste generation step, i.e. during the operation of the reactor. In addition, after remarking the long time span expected from waste generation to their final disposal, the importance of an appropriate record system is pointed out and some basic requirements that should be fulfilled for such system are presented. It is concluded that the tools for a proper characterization of nuclear reactor radioactive waste are available though such tools should be tailored to each specific reactor and their history. (author)

  13. A gas-cooled reactor surface power system

    Science.gov (United States)

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

    1999-01-01

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

  14. Long operation life reactor for lunar surface power

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Dept, University of New Mexico, Albuquerque, NM (United States); Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., University of New Mexico, Albuquerque, NM (United States)

    2011-06-15

    Highlights: > We developed a nuclear reactor with >66 year operation life for surface power. > The reactor, placed below grade and surrounded by regolith, is loaded with fuel pellet on the Moon. > Fuel pellets are launched in separate subcritical canisters. > Examined is the effect of using lunar regolith reflector on the launch mass. - Abstract: The Pellet Bed Reactor (PeBR) with an operational life of 66 full-power years is developed for lunar surface power. It has Inconel X750 structure and vessel and would be launched unfueled then loaded with spherical fuel pellets ({approx}1.0 cm dia.) on the lunar surface after being placed below grade and surrounded with regolith. The pellets, comprised of ZrC-coated UC particles ({approx}850 {mu}m in dia.) dispersed in ZrC matrix, are delivered to the lunar surface in subcritical canisters. The canisters are designed to remain sufficiently subcritical during launch and when submerged in wet sand and flooded with seawater in the unlikely event of a launch abort accident. The PeBR power system nominally generates {approx}100 kW{sub e} at a thermal efficiency of {approx}21% and a reactor exit temperature of 910 K. It employs three separate closed Brayton cycle (CBC) loops each with a turbo-machine unit for energy conversion and two water heat pipes radiator panels for heat rejection. The reactor coolant and CBC working fluid is He-Xe binary gas mixture (40 g/mol). Estimates of the hot-clean excess reactivity and the full-power operation life are obtained using neutronics and fuel depletion analyses. In addition, estimates of the total radioactivity in post-operation PeBR, while being stored below grade on the lunar surface, are determined for up to 1000 years.

  15. Economic risks of nuclear power reactor accidents

    International Nuclear Information System (INIS)

    Models to be used for analyses of economic risks from events which occur during US LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models have been developed for potential use by both the nuclear power industry and regulatory agencies in cost/benefit analyses for decision-making purposes. The new onsite cost models estimate societal losses from power production cost increases, plant capital losses, plant decontamination costs, and plant repair costs which may be incurred after LWR operational events. Early decommissioning costs, plant worker health impact costs, electric utility business costs, nuclear power industry costs, and litigation costs are also addressed. The newly developed offsite economic consequence models estimate The costs of post-accident population protective measures and public health impacts. The costs of population evacuation and temporary relocation, agricultural product disposal, land and property decontamination, and land interdiction are included in the economic models for population protective measures. Costs of health impacts and medical care costs are also included in the models

  16. High density operation for reactor-relevant power exhaust

    Science.gov (United States)

    Wischmeier, M.

    2015-08-01

    With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

  17. Microfluidic reactors for visible-light photocatalytic water purification assisted with thermolysis

    OpenAIRE

    Wang, Ning; Tan, Furui; Wan, Li; Wu, Mengchun; Zhang, Xuming

    2014-01-01

    Photocatalytic water purification using visible light is under intense research in the hope to use sunlight efficiently, but the conventional bulk reactors are slow and complicated. This paper presents an integrated microfluidic planar reactor for visible-light photocatalysis with the merits of fine flow control, short reaction time, small sample volume, and long photocatalyst durability. One additional feature is that it enables one to use both the light and the heat energy of the light sour...

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

    International Nuclear Information System (INIS)

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

  19. Coastdown in light water reactors as a fuel management strategy

    International Nuclear Information System (INIS)

    Improved uranium utilization by means of extended burnup via routine end-of-cycle coastdown has been analyzed, with a specific focus on pressurided water reactors. Both computer and simple analytic models have been developed to determine the optimal coastdown length. Coastdown has been compared with the use of higher fuel-enrichment to achieve comparable burnup values. Temperature and Power coastdown modes were analyzed and changes in the plant thermodynamic efficiency determined. Effects on fuel integrity due to coastdown were examined using a fuel reliability code (SPEAR). Finally the effects on coastdown duration of major parameters involved in charaterizing reactor operation and the economic enviroment were examined. It was found that natural uranium savings up to 7% could be achieved in a typical application by the use of routine pre-planned coast down up to the economic optimun. If coastdown is carried out all the way up to the economic breakeven point yellowcake savings sum up to 16%. Coastdown is substantially more effective than increasing enrichment to extend cycle length without coastdown. Thermodynamic efficiency does not change appreciably during coastdown, a circumstance which greatly simplifies modeling. Coastdown was found to have no statistically significant effect on predicted fuel failure rates. Finally, simple back-of-the evenlope analytic models were found to give an excellent estimate of coastdown duration to both the optimum and breakeven points, and to correctly track the functional behavior induced by all major variables

  20. Emergency action levels for light water reactors. Draft report

    International Nuclear Information System (INIS)

    United States nuclear power plants are currently preparing revised radiological emergency response plans. As part of these plans, each licensee is establishing plant-specific instrumentation readings, called emergency action levels (EALs), that indicate the presence of a problem and the need to declare one of four classes of emergency as designated by the U.S. Nuclear Regulatory Commission. In this report, the EALs prepared for a pressurized water reactor and a boiling water reactor are assessed to determine whether they meet the requirements of NUREG-0654, Rev. 1 Appendix 1. In addition, five recent nuclear incidents are studied to determine how their outcomes might have been affected by the new emergency response plans. Most of the EALs reviewed were judged adequate. In some cases, alternate EALs or sets of generic EALs are proposed to meet the intent of NUREG-0654 Rev 1 Appendix 1, and the wording of some of the initiating conditions for the four emergency classes is clarified. For three of the five incidents reviewed, the new emergency response plans probably would not have affected the outcome; in one case, the plans would probably have prevented core damage; and any effect on the outcome of the last incident is uncertain

  1. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

    This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided

  2. A novel reactor concept for boron neutron capture therapy: annular low-low power reactor (ALLPR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B.; Levine, S.H. [Department of Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States)

    1998-07-01

    Boron Neutron Capture Therapy (BNC), originally proposed in 50's, has been getting renewed attention over the last {approx}10 years. This is in particular due to its potential for treating deep-seated brain tumors by employing epithermal neutron beams. Large (several MW) research reactors are currently used to obtain epithermal beams for BNCT, but because of cost and licensing issues it is not likely that such high-power reactors can be placed in regular medical centers. This paper describes a novel reactor concept for BNCT devised to overcome this obstacle. The design objective was to produce a beam of epithermal neutrons of sufficient intensity for BNCT at <50 kW using low enriched uranium. It is achieved by the annular reactor design, which is called Annular Low-Low Power Reactor (ALLPR). Preliminary studies using Monte Carlo simulations are summarized in this paper. The ALLPR should be relatively economical to build, and safe and easy to operate. This novel concept may increase the viability of using BNCT in medical centers worldwide. (author)

  3. Development and Deployment Strategy for a Small Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    This paper discusses development and deployment strategies for the modular Multi-Application Small Light Water Reactor (MASLWR). Modularity, small size, capability to transport whole modules including containment on road or by rail, simplicity and safety of this reactor allows innovative deployment strategies for a variety of applications. A larger plant may be constructed of many independent power generation units. The multi-module plant is intended to be operated as a base-load plant. Each reactor is to be operated at full load. However, in response to changes in power demand individual units can brought on line or shut down. A larger plant can be built in small increments to match the power demand balancing capital commitments with revenues from sales of electricity. Also, an unplanned shutdown of a reactor only affects a relatively small portion of the total plant capacity. Simplification of MASLWR design and extensive use of modularization coupled with factory fabrication will result in improved productivity of fieldwork and improved quality achieved in a factory environment. The initial MASLWR design concept development has been completed under the U.S. DOE (Department of Energy) Nuclear Energy Research Initiative (NERI) project. This paper discusses a strategy for developing and deploying a MASLWR plant by 2015. This schedule is realistic because the plant design relies on existing industrial experience and manufacturing capabilities. The development strategy consists of the following elements: concept confirmation through testing (under the NERI program a scaled integral test facility has been constructed and initial testing performed), design concept optimization, and design certification based on prototype testing. (authors)

  4. Methods of power reactor decommissioning cost recovery

    International Nuclear Information System (INIS)

    This paper analyzes rate-regulatory tax, accounting and cost recovery factors, and these analyses lead to the following overall conclusions in connection with decommissioning cost recovery. 1) The internal use of accumulated decommissioning funds is strongly recommended because it results in the lowest net ratepayer cost of decommissioning, and 2) The most equitable decommissioning cost recovery method is based on current costs and on the prompt and continuous maintenance of the purchasing power of accumulated funds. Finally, it is noted that the cost recovery approach recommended for decommissioning would have similar advantage if applied to spent fuel cost recovery as well

  5. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  6. Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

    International Nuclear Information System (INIS)

    New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented

  7. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  8. Delayed gamma power measurement for sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Graphical abstract: Display Omitted Research highlights: →20F and 23Ne tagging agents are produced by fast neutron flux. →20F signal has been measured at the SFR Phenix prototype. → A random error of only 3% for an integration time of 2 s could be achieved. →20F and 23Ne power measurement has a reduced temperature influence. → Burn-up impact could be limited by simultaneous 20F and 23Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,α) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  9. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

  10. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  11. Three-dimensional modeling and simulation of vapor explosions in Light Water Reactors

    OpenAIRE

    Schröder, Maxim

    2012-01-01

    Steam explosions can occur during a severe accident in light water nuclear reactors with the core melting as the consequence of interaction of molten core materials with water inside the reactor pressure vessel (in-vessel steam explosions), or after a failure of the reactor vessel due to the release of molten materials into the reactor cavity likely filled with water (ex-vessel steam explosions). Such steam explosions may significantly increase risks of severe accidents threatening the integr...

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

    International Nuclear Information System (INIS)

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

  13. Occupational radiation exposure at commercial nuclear power reactors, 1981. Annual report

    International Nuclear Information System (INIS)

    This report presents an updated compilation of occupational radiation exposures at commercial nuclear power reactors for the years 1969 through 1981. This year's report contains data received from the 70 light water cooled reactors (LWRs) and one high temperature gas cooled reactor that had been declared to be in commercial operation for at least one full year as of December 31, 1981. This represents an increase of two reactors over the number contained in last year's report. The total number of personnel monitored at LWRs in 1981 was 124,504, a slight decrease from that found in 1980. The number of workers that received measurable doses during 1981 was 82,183 which is about 2000 more than that found in 1980. The total collective dose at LWRs for 1981 is estimated to be 54,142 man-rems, which is only about 350 man-rems more than that reported in 1980. The report also presents a summary and some analyses of the exposure data contained in the termination reports that have been submitted by nuclear power licensees to the Commission pursuant to 10 CFR Section 20.408. As of December 31, 1981, personal identification and exposure information had been collected and computerized for some 210,000 of these terminating reactor personnel

  14. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    OpenAIRE

    Kulesza Joel A.; Franceschini Fausto; Evans Thomas M.; Gehin Jess C.

    2016-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a fir...

  15. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  16. Advanced light water reactors: an economically viable part of the world's future energy mix

    International Nuclear Information System (INIS)

    In addition to safety and reliability, a common mission for the international nuclear industry in the 21. century will be ensure affordable electricity. At the Westinghouse Electric Corporation believe our advanced light water reactor (ALWR) design gives us the opportunity to provide the safest, most reliable, lowest cost, most competitive generation method possible for use by nations and utilities worldwide. While the safety and reliability aspects of the ALWR can be proven tangibly and are well-documented, questions have been raised about the technology's ability to work within the world's selling price range for electricity generation. For our industry's financial stability, and especially for the stability of the world's future power needs, Westinghouse has done extensive work on this issue and we are convinced we can meet the competitive challenge. We believe the ALWR can be an economically viable part of the world's future energy mix. This paper will define the competitive challenge that is being addressed by the industry and then analyze three specific areas: capital costs, operating costs, and financing costs. The hidden advantage of nuclear power in responding to these challenges will be explored, and a strong case will be made asserting that the advanced light water reactor will be able to compete in international markets with viable production costs. (authors)

  17. Flow-induced vibration for light water reactors. Progress report, January-June 1980

    International Nuclear Information System (INIS)

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a four-year program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, general scaling laws to improve the accuracy of reduced-scale tests, and the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976, but was suspended on September 30, 1978, due to a shift in Department of Energy (DOE) priorities away from LWR productivity/availability. It was reinitiated as of August 1, 1979. A second program suspension occurred from March 29, 1980 through May 16, 1980, due to funding limits. This progress report summarizes the accomplishments achieved during the period from Janury 1980 to June 1980

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

    International Nuclear Information System (INIS)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. The evaluation of operator reliability factors on power reactor

    International Nuclear Information System (INIS)

    The sophisticated technology system was not assured the reliability system itself because it has contained a part of human dependence affected successfully of reactor operation either how work smoothly and safe or failure ac cured and then accident appears promptly. The evaluation of operator reliability factor on ABWR power reactor has been carried out which consist of criterion skill and workload according to NUREG/CR-2254, NUREG/CR-4016 and NUREG-0835 the reactor operation reliability emphasize to the operator are synergic between skill and workload themselves. The employee's skill will affect to the type and level of their tasks. The operator's skill depend on education and experiences, position or responsibility of tasks, physical conditions (age uninvalid of physic/mental

  1. ELMO Bumpy Torus Reactor and power plant: conceptual design study

    International Nuclear Information System (INIS)

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based on these results, the direction for future improvement on an already attractive reactor design is identified

  2. Light collection and pulse-shape discrimination in elongated scintillator cells for the PROSPECT reactor antineutrino experiment

    Science.gov (United States)

    Ashenfelter, J.; Balantekin, B.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bowes, A.; Brodsky, J. P.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Commeford, K.; Davee, D.; Dean, D.; Deichert, G.; Diwan, M. V.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Goddard, B. W.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Langford, T. J.; Littlejohn, B. R.; Martinez Caicedo, D. A.; McKeown, R. D.; Mendenhall, M. P.; Mueller, P.; Mumm, H. P.; Napolitano, J.; Neilson, R.; Norcini, D.; Pushin, D.; Qian, X.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Sheets, S.; Stemen, N. T.; Surukuchi, P. T.; Varner, R. L.; Viren, B.; Wang, W.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y. R.; Zangakis, G.; Zhang, C.; Zhang, X.

    2015-11-01

    A meter-long, 23-liter EJ-309 liquid scintillator detector has been constructed to study the light collection and pulse-shape discrimination performance of elongated scintillator cells for the PROSPECT reactor antineutrino experiment. The magnitude and uniformity of light collection and neutron-gamma discrimination power in the energy range of antineutrino inverse beta decay products have been studied using gamma and spontaneous fission calibration sources deployed along the cell axis. We also study neutron-gamma discrimination and light collection abilities for differing PMT and reflector configurations. Key design features for optimizing MeV-scale response and background rejection capabilities are identified.

  3. Light Collection and Pulse-Shape Discrimination in Elongated Scintillator Cells for the PROSPECT Reactor Antineutrino Experiment

    CERN Document Server

    Ashenfelter, J; Band, H R; Barclay, G; Bass, C D; Berish, D; Bowden, N S; Bowes, A; Brodsky, J P; Bryan, C D; Cherwinka, J J; Chu, R; Classen, T; Commeford, K; Davee, D; Dean, D; Deichert, G; Diwan, M V; Dolinski, M J; Dolph, J; Dwyer, D A; Gaison, J K; Galindo-Uribarri, A; Gilje, K; Glenn, A; Goddard, B W; Green, M; Han, K; Hans, S; Heeger, K M; Heffron, B; Jaffe, D E; Langford, T J; Littlejohn, B R; Caicedo, D A Martinez; McKeown, R D; Mendenhall, M P; Mueller, P; Mumm, H P; Napolitano, J; Neilson, R; Norcini, D; Pushin, D; Qian, X; Romero, E; Rosero, R; Saldana, L; Seilhan, B S; Sharma, R; Sheets, S; Stemen, N T; Surukuchi, P T; Varner, R L; Viren, B; Wang, W; White, B; White, C; Wilhelmi, J; Williams, C; Wise, T; Yao, H; Yeh, M; Yen, Y -R; Zangakis, G; Zhang, C; Zhang, X

    2015-01-01

    A meter-long, 23-liter EJ-309 liquid scintillator detector has been constructed to study the light collection and pulse-shape discrimination performance of elongated scintillator cells for the PROSPECT reactor antineutrino experiment. The magnitude and uniformity of light collection and neutron/gamma discrimination power in the energy range of antineutrino inverse beta decay products have been studied using gamma and spontaneous fission calibration sources deployed along the cell long axis. We also study neutron-gamma discrimination and light collection abilities for differing PMT and reflector configurations. Key design features for optimizing MeV-scale response and background rejection capabilities are identified.

  4. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

    The assembly known as a CANDU power reactor consists of a number of standardized fuel channels or 'power modules'. Each of these channels produces about 5 thermal megawatts on average. Within practical limitations on fuel enrichment and ultimately on economics, the number of these channels is variable between about 50 and approximately 700. Small reactors suffer from inevitable disadvantages in terms of specific cost of design/construction as well as operating cost. Their natural 'niche' for application is in remote off-grid locations. At the same time this niche application imposes new and strict requirements for staff complement, power system reliability, and so on. The distinct advantage of small reactors arises if the market requires installation of several units in a coordinated installation program - a feature well suited to power requirements in Canada's far North. This paper examines several of the performance requirements and constraints for installation of these plants and presents means for designers to overcome the consequent negative feasibility factors.

  5. Gas-core reactor power transient analysis. Final report

    International Nuclear Information System (INIS)

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

  6. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2012-02-15

    ... COMMISSION Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors AGENCY... ``Decommissioning of Nuclear Power Reactors.'' This guide describes a method NRC considers acceptable for use in... Revision 1 of Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000....

  7. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are...

  8. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  9. Power distribution monitoring and control in the RBMK type reactors

    International Nuclear Information System (INIS)

    Considered are the structures of monitoring and control systems for the RBMK-1000 reactor including three main systems with high independence: the control and safety system (CSS); the system for physical control of energy distribution (SPCED) as well as the Scala system for centralized control (SCC). Main functions and peculiarities of each system are discussed. Main attention is paid to new structural solutions and new equipment components used in these systems. Described are the RBMK operation software and routine of energy distribution control in it. It is noted that the set of reactor control and monitoring systems has a hierarchical structure, the first level of which includes analog systems (CSS and SPCED) normalizing and transmitting detector signals to the systems of the second level based on computers and realizing computer data processing, data representation to the operator, automatic (through CSS) control for energy distribution, diagnostics of equipment condition and local safety with provision for existing reserves with respect to crisis and thermal loading of fuel assemblies. The third level includes a power computer carrying out complex physical and optimization calculations and providing interconnections with the external computer of power system. A typical feature of the complex is the provision of local automatic safety of the reactor from erroneous withdrawal of any control rod. The complex is designed for complete automatization of energy distribution control in reactor in steady and transient operation conditions

  10. Space craft thermal thermionic reactors with flat power distribution

    International Nuclear Information System (INIS)

    The nuclear reactors are potential candidates for energy generation in space missions over longer periods where high power output is required. Among different nuclear energy conversion options, the statical ones, such as thermo-electric or thermionic reactors, are preferable in order to avoid the kinetic disturbances of the space craft and furthermore in order to reduce the failure probabilities to a minimum, caused by lubricants and seals. In the present study, the main parameters of different types of thermal thermionic reactors are discussed which are fueled with U-233 or U-235 and moderated with ZrH1.7 or Beryllium. The investigated thermionic reactors will be layed out to have a constant heat production density on the emitter surface over the space variable, so as to achieve a maximum engineering efficiency with respect to the electrical conversion, nuclear fuel utilization, material damage, thermal and radiation gradients. The power flattening procedure is performed by varying the moderator to fuel ratio, both in axial and radial directions

  11. The production of consuming less: Energy efficiency, climate change, and light bulbs in North Carolina

    Science.gov (United States)

    Thoyre, Autumn

    In this research, I have analyzed the production of consuming less electricity through a case study of promotions of compact fluorescent light bulbs (CFLs). I focused on the CFL because it has been heavily promoted by environmentalists and electricity companies as a key tool for solving climate change, yet such promotions appear counter-intuitive. The magnitude of CFL promotions by environmentalists is surprising because CFLs can only impact less than 1% of U.S. greenhouse gas emissions. CFL promotions by electricity providers are surprising given such companies' normal incentives to sell more of their product. I used political ecological and symbolic interactionist theories, qualitative methods of data collection (including interviews, participant-observation, texts, and images), and a grounded theory analysis to understand this case. My findings suggest that, far from being a self-evident technical entity, energy efficiency is produced as an idea, a part of identities, a resource, and a source of value through social, political, and economic processes. These processes include identity formation and subjectification; gender-coded household labor; and corporate appropriation of household value resulting from environmental governance. I show how environmentalists use CFLs to make and claim neoliberal identities, proposing the concept of green neoliberal identity work as a mechanism through which neoliberal ideologies are translated into practices. I analyze how using this seemingly easy energy efficient technology constitutes labor that is gendered in ways that reflect and reproduce inequalities. I show how electricity companies have used environmental governance to valorize and appropriate home energy efficiency as an accumulation strategy. I conclude by discussing the symbolic power of CFLs, proposing a theory of green obsolescence, and framing the production of energy efficiency as a global production network. I found that promoting energy efficiency involves

  12. Accident Tolerant Fuel Concepts for Light Water Reactors. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    Nuclear fuel is a highly complex material that has been subject to continuous development over the past 40 years and has reached a stage where it can be safely and reliably irradiated up to 65 GWd/tU in commercial nuclear reactors. During this time, there have been many improvements to the original designs and materials used. However, the basic design of uranium oxide fuel pellets clad with zirconium alloy tubing has remained the fuel choice for the vast majority of commercial nuclear power plants. Severe accidents, such as those at the Three Mile Island and Fukushima Daiichi have shown that under such extreme conditions, nuclear fuel will fail and the high temperature reactions between zirconoi alloys and water will lead to the generation of hydrogen, with the potential for explosions to occur, daming the plant further. Recognizing that the current fuel designs are vulnerable to severe accident conditions, tehre is renewed interesst in alternative fuel designs that would be more resistant to fuel failure and hydrogen production. Such new fuel designs will need to be compatible with existing fuel and reactor systems if they are to be utilized in the current reactor fleet and in current new build designs, but there is also the possibility of new designs for new reactor systems. This publication provides a record of the Technical Meeting on Accident Tolerant Fuel Concepts for Light Water Reactors, held at Oak Ridge National Laboratories (ORNL), United States of America, 13-16 October 2014, to consider the early stages of research and development into accident tolerant fuel. There were 45 participants from 10 countries taking part in the meeting, with 32 papers organized into 7 sessions, of which 27 are included in this publication. This meeting is part of a wider investigation into such designs, and it is anticipated that further Technical Meetings and research programmes will be undertaken in this field

  13. SUPERNOVA LIGHT CURVES POWERED BY FALLBACK ACCRETION

    Energy Technology Data Exchange (ETDEWEB)

    Dexter, Jason; Kasen, Daniel, E-mail: jdexter@berkeley.edu [Departments of Physics and Astronomy, University of California, Berkeley, CA 94720 (United States)

    2013-07-20

    Some fraction of the material ejected in a core collapse supernova explosion may remain bound to the compact remnant, and eventually turn around and fall back. We show that the late time ({approx}>days) power potentially associated with the accretion of this 'fallback' material could significantly affect the optical light curve, in some cases producing super-luminous or otherwise peculiar supernovae. We use spherically symmetric hydrodynamical models to estimate the accretion rate at late times for a range of progenitor masses and radii and explosion energies. The accretion rate onto the proto-neutron star or black hole decreases as M-dot {proportional_to}t{sup -5/3} at late times, but its normalization can be significantly enhanced at low explosion energies, in very massive stars, or if a strong reverse shock wave forms at the helium/hydrogen interface in the progenitor. If the resulting super-Eddington accretion drives an outflow which thermalizes in the outgoing ejecta, the supernova debris will be re-energized at a time when photons can diffuse out efficiently. The resulting light curves are different and more diverse than previous fallback supernova models which ignored the input of accretion power and produced short-lived, dim transients. The possible outcomes when fallback accretion power is significant include super-luminous ({approx}> 10{sup 44} erg s{sup -1}) Type II events of both short and long durations, as well as luminous Type I events from compact stars that may have experienced significant mass loss. Accretion power may unbind the remaining infalling material, causing a sudden decrease in the brightness of some long duration Type II events. This scenario may be relevant for explaining some of the recently discovered classes of peculiar and rare supernovae.

  14. Supernova Light Curves Powered by Fallback Accretion

    Science.gov (United States)

    Dexter, Jason; Kasen, Daniel

    2013-07-01

    Some fraction of the material ejected in a core collapse supernova explosion may remain bound to the compact remnant, and eventually turn around and fall back. We show that the late time (gsimdays) power potentially associated with the accretion of this "fallback" material could significantly affect the optical light curve, in some cases producing super-luminous or otherwise peculiar supernovae. We use spherically symmetric hydrodynamical models to estimate the accretion rate at late times for a range of progenitor masses and radii and explosion energies. The accretion rate onto the proto-neutron star or black hole decreases as \\dot{M} \\propto t^{-5/3} at late times, but its normalization can be significantly enhanced at low explosion energies, in very massive stars, or if a strong reverse shock wave forms at the helium/hydrogen interface in the progenitor. If the resulting super-Eddington accretion drives an outflow which thermalizes in the outgoing ejecta, the supernova debris will be re-energized at a time when photons can diffuse out efficiently. The resulting light curves are different and more diverse than previous fallback supernova models which ignored the input of accretion power and produced short-lived, dim transients. The possible outcomes when fallback accretion power is significant include super-luminous (gsim 1044 erg s-1) Type II events of both short and long durations, as well as luminous Type I events from compact stars that may have experienced significant mass loss. Accretion power may unbind the remaining infalling material, causing a sudden decrease in the brightness of some long duration Type II events. This scenario may be relevant for explaining some of the recently discovered classes of peculiar and rare supernovae.

  15. Reactor Power for Large Displacement Autonomous Underwater Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    McClure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-24

    This is a PentaChart on reactor power for large displacement autonomous underwater vehicles. Currently AUVs use batteries or combinations of batteries and fuel cells for power. Battery/fuel cell technology is limited by duration. Batteries and cell fuels are a good match for some missions, but other missions could benefit greatly by a longer duration. The goal is the following: to design nuclear systems to power an AUV and meet design constraints including non-proliferation issues, power level, size constraints, and power conversion limitations. The action plan is to continue development of a range of systems for terrestrial systems and focus on a system for Titan Moon as alternative to Pu-238 for NASA.

  16. Imperatives for using plutonium in commercial power reactors

    International Nuclear Information System (INIS)

    The use of reprocessed or newly produced plutonium as a fissile fuel in commercial nuclear reactors in the US has been actively suppressed by the current US Administration. Yet, many other advanced nations have already adopted mixed oxide fuels which are manufactured from a mixture of plutonium and natural uranium compounds. These nations have successfully proven the use of such nuclear fuel in their commercial power reactors for many years. The full consequence of the restrictive nuclear policy in the US will greatly limit the lifetime of the nuclear fuel resources in the US from a nominal potential of 100 centuries or more of potential energy supply to about 50 years or less at economical prices for uranium. This paper addresses both the imperatives and the potential and the perceived hazards of plutonium utilization and examines the consequences of government policy regarding utilization of nuclear power

  17. Characteristics of a reactor with power reactivity feedback

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    The point-reactor model with power reactivity feedback becomes a nonlinear system. Its dynamic characteristic shows great complexity. According to the mathematic definition of stability in differential equa-tion qualitative theory, the model of a reactor with power reactivity feedback is judged unstable. The equilib-rium point is a saddle-node point. A portion of the trajectory in the neighborhood of the equilibrium point is parabolic fan curve, and the other is hyperbolic fan curve. Based on phase locus near the equilibrium point, it is pointed out that the model is still stable within physical limits. The difference between stabilities in the mathematical sense and in the physical sense is indicated.

  18. Technological implications of SNAP reactor power system development on future space nuclear power systems

    International Nuclear Information System (INIS)

    Nuclear reactor systems are one method of satisfying space mission power needs. The development of such systems must proceed on a path consistent with mission needs and schedules. This path, or technology roadmap, starts from the power system technology data base available today. Much of this data base was established during the 1960s and early 1970s, when government and industry developed space nuclear reactor systems for steady-state power and propulsion. One of the largest development programs was the Systems for Nuclear Auxiliary Power (SNAP) Program. By the early 1970s, a technology base had evolved from this program at the system, subsystem, and component levels. There are many implications of this technology base on future reactor power systems. A review of this base highlights the need for performing a power system technology and mission overview study. Such a study is currently being performed by Rockwell's Energy Systems Group for the Department of Energy and will assess power system capabilities versus mission needs, considering development, schedule, and cost implications. The end product of the study will be a technology roadmap to guide reactor power system development

  19. Circuit for power variation rate measurements in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moisin, L.H.

    1980-01-01

    An asychronous digital circuit for the power variation rate of a nuclear reactor is proposed. The circuit is based on the fact that the variation rate can be obtained by a simple division between the difference of two time normalized adjacent measurements of the neutron flux and the total value of the first measurement. The circuit maintains a constant precision of the counting rate due to the effect of an automatic time constant switch. 4 references.

  20. Summary of space nuclear reactor power systems, 1983--1992

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  1. Dynamic power behavior of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.)

  2. Summary of space nuclear reactor power systems, 1983--1992

    International Nuclear Information System (INIS)

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power

  3. 75 FR 77919 - Carolina Power & Light Company Shearon Harris Nuclear Power Plant, Unit 1; Environmental...

    Science.gov (United States)

    2010-12-14

    ... part 73 as discussed in a Federal Register notice dated March 27, 2009 (74 FR 13926). There will be no... dedicated safe shut down (SSD) equipment of HNP, Unit 1 within the room in which some important security... available due to design basis tornado and missile considerations for the safety related conduits and...

  4. Autonomous Control Capabilities for Space Reactor Power Systems

    Science.gov (United States)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  5. Economic evaluation of nuclear reactor operation utilizing power effect

    International Nuclear Information System (INIS)

    The operation of a reactor at the so-called power effect may substantially increase the burnup of fuel to be removed. The aim of the evaluation of such reactor operation is the optimal determination of the time over which the yield of the higher use of fuel exceeds economic losses resulting from the increased share of constant expenditure of the price of generated kWh of electric power which ensues from such operation. A mathematical model is presented for such evaluation of reactor operation with regard to benefits for the national economy which is the basis of the ESTER 2 computer program. The calculations show that the prices of generated and delivered kWh are minimally 2% less than the prices of generated power without the power effect use. The minimum ranges in the interval of 30 to 50 days. The dependence of the price of generated and delivered kWh from the point of view of the operator of the power plant as well as the component of fuel price of generated kWh will not reach the minimum even after 50 days of operation. From the operating and physical points of view the duration of power effect is not expected to exceed 20 to 30 days which means that from the point of view of the national economy the price of generated and delivered kWh will be 1.6 to 2% less and the fuel component of the price of the generated kWh will be 3 to 4.5% lower. (Z.M.). 5 figs., 3 refs

  6. Investigation of activity release during light water reactor core meltdown

    International Nuclear Information System (INIS)

    A test facility was developed for the determination of activity release and of aerosol characteristics under realistic light water reactor core melting conditions. It is composed of a high-frequency induction furnace, a ThO2 crucible system, and a collection apparatus consisting of membrane and particulate filters. Thirty-gram samples of a representative core material mixture (corium) were melted under air, argon, or steam at 0.8 to 2.2 bar. In air at 27000C, for example, the relative release was 0.4 to 0.7% for iron, chromium, and cobalt and 4 to 11% for tin, antimony, and manganese. Higher release values of 20 to 40% at lower temperatures (21500C, air) were found for selenium, cadmium, tellurium, and cesium. The size distribution of the aerosol particles was trimodal with maxima at diameters of 0.17, 0.30, and 0.73 μm. The result of a qualitative x-ray microanalysis was that the main elements of the melt were contained in each aerosol particle. Further investigations will include larger melt masses and the additional influence of concrete on the release and aerosol behavior

  7. Light water reactor pressure isolation valve performance testing

    International Nuclear Information System (INIS)

    The Light Water Reactor Valve Performance Testing Program was initiated by the NRC to evaluate leakage as an indication of valve condition, provide input to Section XI of the ASME Code, evaluate emission monitoring for condition and degradation and in-service inspection techniques. Six typical check and gate valves were purchased for testing at typical plant conditions (550F at 2250 psig) for an assumed number of cycles for a 40-year plant lifetime. Tests revealed that there were variances between the test results and the present statement of the Code; however, the testing was not conclusive. The life cycle tests showed that high tech acoustic emission can be utilized to trend small leaks, that specific motor signature measurement on gate valves can trend and indicate potential failure, and that in-service inspection techniques for check valves was shown to be both feasible and an excellent preventive maintenance indicator. Life cycle testing performed here did not cause large valve leakage typical of some plant operation. Other testing is required to fully understand the implication of these results and the required program to fully implement them. (author)

  8. Studies of severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    From 10 to 12 November 1986 some 80 delegates met under the auspices of the CEC working group on the safety of light-water reactors. The participants from EC Member States were joined by colleagues from Sweden, Finland and the USA and met to discuss the subject of severe accidents in LWRs. Although this seminar had been planned well before Chernobyl, the ''severe-accident-that-really-happened'' made its mark on the seminar. The four main seminar topics were: (i) high source-term accident sequences identified in PSAs, (ii) containment performance, (iii) mitigation of core melt consequences, (iv) severe accident management in LWRs. In addition to the final panel discussion there was also a separate panel discussion on lessons learned from the Chernobyl accident. These proceedings include the papers presented during the seminar and they are arranged following the seminar programme outline. The presentations and discussions of the two panels are not included in the proceedings. The general conclusions and directions following from these two panels were, however, considered in a seminar review paper which was published in the March 1987 issue of Nuclear Engineering International

  9. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    Energy Technology Data Exchange (ETDEWEB)

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P. [Division of Applied Nuclear Physics, Uppsala University, Aangstroemlaboratoriet Laegerhyddsvaegen 1, 751 20 Uppsala (Sweden)

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  10. Environmental impacts of nonfusion power systems. [Data on environmental effects of all power sources that may be competitive with fusion reactor power plants

    Energy Technology Data Exchange (ETDEWEB)

    Brouns, R.J.

    1976-09-01

    Data were collected on the environmental effects of power sources that may be competitive with future fusion reactor power plants. Data are included on nuclear power plants using HTGR, LMBR, GCFR, LMFBR, and molten salt reactors; fossil-fuel electric power plants; geothermal power plants; solar energy power plants, including satellite-based solar systems; wind energy power plants; ocean thermal gradient power plants; tidal energy power plants; and power plants using hydrogen and other synthetic fuels as energy sources.

  11. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Paul Y [Los Alamos National Laboratory

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  12. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Science.gov (United States)

    2013-07-10

    ... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...- Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area includes a revised......

  13. LIGHT WATER REACTOR SUSTAINABILITY PROGRAM ADVANCED INSTRUMENTATION, INFORMATION, AND CONTROL SYSTEMS TECHNOLOGIES TECHNICAL PROGRAM PLAN FOR 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce; Thomas, Ken

    2014-07-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  14. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2014-09-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  15. Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

    International Nuclear Information System (INIS)

    Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to execute

  16. New designs of medium power WWER reactor plants

    International Nuclear Information System (INIS)

    The task of constructing NPPs as the objects of regional power industry is included into the Federal Target Program on nuclear power technologies of new generation for the period till 2020. Such NPPs are considered as perspective sources of energy for solution of the problems concerning provision of electric energy, household and industrial heat to the regions with limited capabilities of the power grid. OKB 'GIDROPRESS' present the conceptual study of RP design for the Unit of 600 MW (el.) power, taking into account their long-term experience in the field of development and operation of WWER reactor plants. Practical implementation of WWER-600 and WWER-300 RP designs seems to be feasible: practice in manufacturing the main equipment is available; cooperation of design, scientific organizations and manufacturers of equipment; is established; basic design solutions for equipment are of reference character

  17. Reactor Power for Large Displacement Autonomous Underwater Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    McClure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-24

    This is a PentaChart on reactor power for large displacement autonomous underwater vehicles. Currently AUVs use batteries or combinations of batteries and fuel cells for power. Battery/fuel cell technology is limited by duration. Batteries and cell fuels are a good match for some missions, but other missions could benefit greatly by a longer duration. The goal is the following: to design nuclear systems to powr an AUV and meet design constraints including non-proliferation issues, power level, size constraints, and power conversion limitations. The action plan is to continue development of a range of systems for terrestrial systems and focus on a system for TItan Moon as alternative to Pu-238 for NASA.

  18. Supercritical Water Reactor Cycle for Medium Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  19. Power Excursion Accident Analysis of Research Water Reactor

    International Nuclear Information System (INIS)

    A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

  20. Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

    International Nuclear Information System (INIS)

    In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R and D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R and D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10-12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I and C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy's Light Water Reactor Sustainability Program. DOE

  1. Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Baldwin; Magdy Tawfik; Leonard Bond

    2010-06-01

    In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has

  2. Proliferation resistant fuel cycle system for the transition from light water reactors to fast reactors

    International Nuclear Information System (INIS)

    Full text: Introduction of commercial fast reactors (FR) is predicted to start around 2050 in Japan. Effective utilization of plutonium in FR is important for the sustainable electricity generation by nuclear. Successive replacement of light water reactors (LWR) to FR will take more than 60-years and reasonable fuel cycle management is necessary during this period. The transition scenario has various unpredictable factors such as introduction speed and time of FR, and flexible fuel cycle system was proposed to respond to these factors. The system consists of LWR and FR spent fuels reprocessing for reduction of LWR spent fuel volume and FR fuel fabrication. LWR fuel reprocessing only carries out about 90% uranium removal from LWR spent fuel, then the composition of residual spent fuel called recycle material is about 50% uranium, 15% plutonium and 35% fission products + minor actinides. Recycle material is transferred to FR fuel reprocessing to recover plutonium and uranium followed by mixed oxide (MOX) fuel fabrication for FR with radioactive impurities. Depending on the introduction time of FR, recycle material (about 1/10 volume of original spent fuel) may be stored for future use. The system has some characteristics compared with ordinary system that consists of full reprocessing facilities for LWR and FR spent fuels to produce FR fresh fuels. The LWR reprocessing facility becomes much smaller due to no Pu-U recovery and fabrication. The recycle materials can supply higher content of plutonium to FR and be compactly stored in case of FR introduction delay. Plutonium always contains uranium and impurities (fission products and minor actinides), thus the system maintains high proliferation resistance. The plutonium balance was calculated under several conditions, which revealed that the system could supply enough and no excess plutonium to FR

  3. Design Concept for a Nuclear Reactor-Powered Mars Rover

    Science.gov (United States)

    Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

    2003-01-01

    A study was recently carried out by a team from JPL and the DOE to investigate the utility of a DOE-developed 3 kWe surface fission power system for Mars missions. The team was originally tasked to perform a study to evaluate the usefulness and feasibility of incorporation of such a power system into a landed mission. In the course of the study it became clear that the application of such a power system was enabling to a wide variety of potential missions. Of these, two missions were developed, one for a stationary lander and one for a reactor-powered rover. This paper discusses the design of the rover mission, which was developed around the concept of incorporating the fission power system directly into a large rover chassis to provide high power, long range traverse capability. The rover design is based on a minimum extrapolation of technology, and adapts existing concepts developed at JPL for the 2009 Mars Science Laboratory (MSL) rover, lander and EDL systems. The small size of the reactor allowed its incorporation directly into an existing large MSL rover chassis design, allowing direct use of MSL aeroshell and pallet lander elements, beefed up to support the significantly greater mass involved in the nuclear power system and its associated shielding. This paper describes the unique design challenges encountered in the development of this mission architecture and incorporation of the fission power system in the rover, and presents a detailed description of the final design of this innovative concept for providing long range, long duration mobility on Mars.

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

    International Nuclear Information System (INIS)

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

  5. Evaluation of the qualification of SPERT [Special Power Excursion Reactor Test] fuel for use in non-power reactors

    International Nuclear Information System (INIS)

    This report summarizes the US Nuclear Regulatory Commission staff's evaluation of the qualification of the stainless-steel-clad uranium/oxide (UO2) fuel pins for use in non-power reactors. The fuel pins were originally procured in the 1960's as part of the Special Power Excursion Reactor Test (SPERT) program. Argonne National Laboratory (ANL) examined 600 SPERT fuel pins to verify that the pins were produced according to specification and to assess their present condition. The pins were visually inspected under 6X magnification and by X-radiographic, destructive, and metallographic examinations. Spectrographic and chemical analyses were performed on the UO2 fuel. The results of the qualification examinations indicated that the SPERT fuel pins meet the requirements of Phillips Specification No. F-1-SPT and have suffered no physical damage since fabrication. Therefore, the qualification results give reasonable assurance that the SPERT fuel rods are suitable for use in non-power reactors provided that the effects of thin-wall defects in the region of the upper end cap and low-density fuel pellets are evaluated for the intended operating conditions. 1 ref., 4 figs., 11 tabs

  6. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  7. Light Water Reactor Sustainability Research and Development Program Plan. Fiscal Year 2009-2013

    International Nuclear Information System (INIS)

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R and D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R and D programs. The purpose of the LWRS R and D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R and D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its

  8. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    Energy Technology Data Exchange (ETDEWEB)

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement

  9. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept

    Directory of Open Access Journals (Sweden)

    Haileyesus Tsige-Tamirat

    2015-01-01

    Full Text Available Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required.

  10. Operational margin monitoring system for boiling water reactor power plants

    International Nuclear Information System (INIS)

    This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

  11. Temperature effects studies in light water reactor lattices

    International Nuclear Information System (INIS)

    The CREOLE experiments performed in the EOLE critical facility located in the Nuclear Center of CADARACHE - CEA (UO2 and UO2-PuO2 lattice reactivity temperature coefficient continuous measurements between 200C and 3000C; integral measurements by boron equivalent effect in the moderator; water density effects measurements with the use of over cladding aluminium tubes to remove moderator) allow to get an interesting and complete information on the temperature effects in the light water reactor lattices. A very elaborated calcurated scheme using the transport theory and the APOLLO cross sections library, has been developed. The analysed results of the whole lot of experiments show that the discrepancy between theory and experiment strongly depends on the temperature range and on the type of lattices considered. The error is mainly linked with the thermal spectrum effects. A study on the temperature coefficient sensitivity to the different cell neutron parameters has shown that only the shapes of the 235U and 238U thermal cross sections have enough weight and uncertainty margins to explain the observed experimental/calculation bias. Instead of arbitrarily fitting the identified wrong data on the calculation of the reactivity temperature coefficient we have defined a procedure of modification of the cross sections based on the consideration of the basic nuclear data: resonance parameters and associated statistic laws. The implementation of this procedure has led to propose new thermal cross sections sets for 235U and 238U consistent with the uncertainty margins associated with the previously accepted values and with some experimental data

  12. Study of plutonium recycling physics in light water reactors

    International Nuclear Information System (INIS)

    A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters

  13. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

  14. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 2

    International Nuclear Information System (INIS)

    This environmental statement assesses the impacts of the implementation of plutonium recycle in the LWR industry. It is based on assumptions that are intended to reflect conservatively an acceptable level of the application of current technology. It is not intended to be a representation of the ''as low as reasonably achievable'' (ALARA) philosophy. This generic environmental statement discusses the anticipated effects of recycling plutonium in light water nuclear power reactors. It is based on about 30 years of experience with the element in the context of a projected light water nuclear power industry that is already substantial. A background perspective on plutonium, its safety, and its recycling as a reactor fuel is presented

  15. Axial offset as measure of stability of light water nuclear reactor during capacity maneuvering

    Directory of Open Access Journals (Sweden)

    Mark V. Nikolsky

    2015-03-01

    Full Text Available High reliability and security of power unit are required during operation of power unit while maneuvering. They depend on the stability of reactor when transition from one power level to another. The axial offset is a quantitative measure of the reactor stability. It is shown that change of the active core inlet coolant temperature yields an uncontrollable disturbance affecting the axial offset and therefore the reactor stability. To insure the reactor stability the compromise-combined power control method is proposed. Analysis of the influence of temperature of coolant at the magnitude of the axial offset for different regulatory programs is carried out. The change in the depth of immersion of regulators in the active zone for different regulatory programs when the reactor plant daily capacity maneuver is studied.

  16. LWR-WIMS, a computer code for light water reactor lattice calculations

    International Nuclear Information System (INIS)

    LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)

  17. Catalogue and classification of technical safety rules for light-water reactors and reprocessing plants

    International Nuclear Information System (INIS)

    This report on the cataloguing and classification of technical rules for land-based light-water reactors and reprocessing plants contains a list of classified rules. The reasons for the classification system used are given and discussed

  18. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  19. Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

    International Nuclear Information System (INIS)

    Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

  20. SWR 1000: The new boiling water reactor power plant concept

    International Nuclear Information System (INIS)

    Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

  1. Source-term reevaluation for US commercial nuclear power reactors: a status report

    Energy Technology Data Exchange (ETDEWEB)

    Herzenberg, C.L.; Ball, J.R.; Ramaswami, D.

    1984-12-01

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date.

  2. Source-term reevaluation for US commercial nuclear power reactors: a status report

    International Nuclear Information System (INIS)

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date

  3. Categorization of failed and damaged spent LWR [light-water reactor] fuel currently in storage

    International Nuclear Information System (INIS)

    The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs

  4. Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

    International Nuclear Information System (INIS)

    The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security

  5. ENFORM II: a calculational system for light water reactor logistics and effluent analysis

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Lewallen, M.A.; Purcell, W.L.; Cole, B.M.

    1979-09-01

    ENFORM is a computer-based information system that addresses the material logistics, environmental releases and economics of light water reactor (LWR) operation. The most important system inputs consist of electric energy generation requirements, details of plant construction scheduling, unit costs, and environmental release factors. From these inputs the ENFORM system computes the mass balances and generates the environmental release information for noxious chemicals and radionuclides from various fuel cycle facilities (except waste disposal). Fuel cycle costs and electric power costs are also computed. All code development subsequent to 1977 is summarized. Programming instructions are provided for the modules that are comprised in the ENFORM system. ENGEN, a code that uses a generation schedule specified by the user and isotopic data generated by ORIGEN, has been developed to produce a scenario-specific data base. Other codes (ENMAT, ENRAD, etc) have been developed to use data base information to estimate radioactive and nonradioactive release information.

  6. ENFORM II: a calculational system for light water reactor logistics and effluent analysis

    International Nuclear Information System (INIS)

    ENFORM is a computer-based information system that addresses the material logistics, environmental releases and economics of light water reactor (LWR) operation. The most important system inputs consist of electric energy generation requirements, details of plant construction scheduling, unit costs, and environmental release factors. From these inputs the ENFORM system computes the mass balances and generates the environmental release information for noxious chemicals and radionuclides from various fuel cycle facilities (except waste disposal). Fuel cycle costs and electric power costs are also computed. All code development subsequent to 1977 is summarized. Programming instructions are provided for the modules that are comprised in the ENFORM system. ENGEN, a code that uses a generation schedule specified by the user and isotopic data generated by ORIGEN, has been developed to produce a scenario-specific data base. Other codes (ENMAT, ENRAD, etc) have been developed to use data base information to estimate radioactive and nonradioactive release information

  7. Fission product release in accidents in light water reactors

    International Nuclear Information System (INIS)

    The author deals with the three phases of release from the reactor core, from the reactor system, and finally from the containment. Particular interest is given to the release from the reactor core at temperatures which let the fuel rod cladding burst leading to meltdown of the fuel elements and evaporation from the core melt. The special case of the steam explosion with small nuclear fuel particles pouring out into an oxidating atmosphere is touched upon. The Rasmussen study is the basis of the statements. (orig./LH)

  8. Small space reactor power systems for unmanned solar system exploration missions

    International Nuclear Information System (INIS)

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model

  9. Demographic statistics pertaining to nuclear power reactor sites

    International Nuclear Information System (INIS)

    Population statistics are presented for 145 nuclear power plant sites. Summary tables and figures are included that were developed to aid in the evaluation of trends and general patterns associated with the various parameters of interest, such as the proximity of nuclear plant sites to centers of population. The primary reason for publishing this information at this time is to provide a factual basis for use in discussions on the subject of reactor siting policy. The report is a revised and updated version of a draft report published in December 1977. Errors in the population data base have been corrected and new data tabulations added

  10. HAZARDS SUMMARY REPORT FOR THE ARMY PACKAGE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1955-07-27

    The APPR-I is described and the various hazards are reviewed. Because of the reactor's location near the nation's Capitol, containment is of the utmost importance. The maximum energy release in any possible accident is 7.4 million Btu's which is completely contained within a 7/8 inch thick steel cylindrical shell with hemispherical ends. The vapor container is 60 ft high and 32 ft in diameter and is lined on the inside with 2 ft of reinforced concrete which provides missile protection and is part of the secondary shield. All possible nuclear excursions are reviewed and the energy from any of these is insignificant compared to the stored energy in the water. The maximum credible accident is caused hy the reactor running constantly at its maximum power of 10 Mw and through an extremely unlikely sequence of failures, causing the temperature of the water in the primary and secondary systeras to rise to saturation; whereupon a rupture occurs releasing the stored energy of 7.4 million Btu's into the vapor container. If the reactor core melts during the incident, a maximum of 10/sup 8/ curies of activity is released. While it appears impossible for a rupture of the vapor container to oecur except by sabotage or bombing, the hazards to the surrounding area are discussed in the event of such a rupture occurring simultaneously with the maximum credible accident. (auth)

  11. Neutron dose estimation in a zero power nuclear reactor

    Science.gov (United States)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

    2016-10-01

    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  12. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Science.gov (United States)

    2010-01-01

    ... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked...-Cooled Power Reactors J Appendix J to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. J Appendix J to Part 50—Primary Reactor...

  13. Utilization of Minor Actinides (Np, Am, Cm) in Nuclear Power Reactor

    Science.gov (United States)

    Gerasimov, A.; Bergelson, B.; Tikhomirov, G.

    2014-06-01

    Calculation research of the utilization process of minor actinides (transmutation with use of power released) is performed for specialized power reactor of the VVER type operating on the level of electric power of 1000 MW. Five subsequent cycles are considered for the reactor with fuel elements containing minor actinides along with enriched uranium. It was shown that one specialized reactor for the one cycle (900 days) can utilize minor actinides from several VVER-1000 reactors without any technological and structural modifications. Power released because of minor actinide fission is about 4% with respect to the total power

  14. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    Science.gov (United States)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  15. Role of VVER-type reactors in large-scale nuclear power of the XXI century

    International Nuclear Information System (INIS)

    Light water reactors (LWR) make over 85% of the world nuclear park and are presently constructed in 12 countries. One of the generally recognized LWR development directions is represented by VVER reactor concept, created and developed in the former Soviet Union. For over 35 years the VVER existence (with gross capacities ranging from 70 to 100 MWe), 58 power units have been built, and 49 are still in operation (13 in Russia and Ukraine each, 6 - in Bulgaria and Slovakia each, 4 - in Hungary and Czech Republic each, 2 - in Finland and 1 - in Armenia). The oldest of operating VVERs -unit 3 of Novovoronezh NPP in Russia - was connected to grid in 1971; the last - Mochovce-2 in Slovakia - was launched in 1999. Geography of VVER reactors is developing quite dynamically. For the first time this reactor type is being built in the countries of Asia: China and Iran, as well as in Cuba. Construction of the first VVER in India is also expected. (author)

  16. A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Highlights: • Data analysis for high-performance simulations of reactors will be a problem that we address with a new management system. • We describe new input-output libraries for nuclear reactor simulations. • We describe a new user interface for visualizing and analyzing simulation results. • We show the utility of these systems with a 17 × 17 fuel assembly example simulation. • The availability of the code and avenues for collaboration are presented. - Abstract: Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the system’s requirements and design

  17. Screening action potentials: The power of light

    Directory of Open Access Journals (Sweden)

    Lars eKaestner

    2011-07-01

    Full Text Available Action potentials reflect the concerted activity of all electrogenic constituents in the plasma membrane during the excitation of a cell. Therefore, the action potential is an integrated readout and a promising parameter to detect electrophysiological failures or modifications thereof in diagnosis as well as in drug screens. Cellular action potentials can be recorded by optical approaches. To fulfill the pre-requirements to scale up for e.g. pharmacological screens the following preparatory work has to be provided: (i model cells under investigation need to represent target cells in the best possible manner; (ii optical sensors that can be either small molecule dyes or genetically encoded potential probes need to provide a reliable readout with minimal interaction with the naive behavior of the cells and (iii devices need to be capable to stimulate the cells, read out the signals with the appropriate speed as well as provide the capacity for a sufficient throughput. Here we discuss several scenarios for all three categories in the field of cardiac physiology and pharmacology and provide a perspective to use the power of light in screening cardiac action potentials.

  18. Supernova Light Curves Powered by Fallback Accretion

    CERN Document Server

    Dexter, Jason

    2012-01-01

    Some fraction of the material ejected in a core collapse supernova explosion may remain bound to the compact remnant, and eventually turn around and fall back. We show that the late time (> days) power associated with the accretion of this "fallback" material may significantly affect the optical light curve, in some cases producing super-luminous or otherwise peculiar supernovae. We use spherically symmetric hydrodynamical models to estimate the accretion rate at late times for a range of progenitor masses and radii and explosion energies. The accretion rate onto the proto-neutron star or black hole decreases as Mdot ~ t^-5/3 at late times, but its normalization can be significantly enhanced at low explosion energies, in very massive stars, or if a strong reverse shock wave forms at the helium/hydrogen interface in the progenitor. If the resulting super-Eddington accretion drives an outflow which thermalizes in the outgoing ejecta, the supernova debris will be re-energized at a time when photons can diffuse o...

  19. Microfluidic reactors for visible-light photocatalytic water purification assisted with thermolysis.

    Science.gov (United States)

    Wang, Ning; Tan, Furui; Wan, Li; Wu, Mengchun; Zhang, Xuming

    2014-09-01

    Photocatalytic water purification using visible light is under intense research in the hope to use sunlight efficiently, but the conventional bulk reactors are slow and complicated. This paper presents an integrated microfluidic planar reactor for visible-light photocatalysis with the merits of fine flow control, short reaction time, small sample volume, and long photocatalyst durability. One additional feature is that it enables one to use both the light and the heat energy of the light source simultaneously. The reactor consists of a BiVO4-coated glass as the substrate, a blank glass slide as the cover, and a UV-curable adhesive layer as the spacer and sealant. A blue light emitting diode panel (footprint 10 mm × 10 mm) is mounted on the microreactor to provide uniform irradiation over the whole reactor chamber, ensuring optimal utilization of the photons and easy adjustments of the light intensity and the reaction temperature. This microreactor may provide a versatile platform for studying the photocatalysis under combined conditions such as different temperatures, different light intensities, and different flow rates. Moreover, the microreactor demonstrates significant photodegradation with a reaction time of about 10 s, much shorter than typically a few hours using the bulk reactors, showing its potential as a rapid kit for characterization of photocatalyst performance. PMID:25584117

  20. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)

    1997-12-31

    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  1. Light Water Reactor LOCA Safety Parameter Analysis Using The Reactor Dynamics Code Dynode and Subchanflow

    International Nuclear Information System (INIS)

    Safety margin evaluation for LWR can be strongly improved by means of multi-physics and multi-scale methodologies. Thus, projects like NURISP have been focused on incorporating into one computational platform the latest advances in reactor simulation tools. An important task of KIT in the frame of NURISP was the improvement, extension and integration of the pin power reconstruction method (PPR) of DYN3D. The flexibility of the new development allows LOCA refinement in the spatial mesh for specific regions of interest or even having a whole core with pin-by-pin resolution. This integration is a step forward in the direction of two-level coupling with the subchannel code FLICA, being one of the major objectives of NURISP. In order to investigate the scope of the new DYN3D implementation, an off-line fast running methodology for the evaluation of LOCA safety parameters using DYN3D and the subchannel code SUBCHANFLOW has been developed. The DYN3D-PPR method has been used to obtain a time-dependent 3D pin power distribution in the hottest assembly of a mini-core during a transient. The bundle power maps obtained in this way are used as an input for SUBCHANFLOW to evaluate the time-dependent variation of LOCA safety parameters such as DNBR or the maximum cladding temperature. As a result, a 3D-map of LOCA safety parameters is available for each simulation time step. On the other hand, the stand-alone version of DYN3D contains an internally coupled 1D thermal-hydraulic model LOCA that provides the on-line thermal-hydraulic feedback for the updating of cross sections. Additionally it allows also evaluating LOCA safety parameters using the new methodology and DYN3D stand-alone calculation is presented. As expected, the use of SUBCHANFLOW yields a more detailed and accurate prediction of LOCA safety parameters comparing with the 1D thermal-hydraulics model LOCA

  2. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    Science.gov (United States)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  3. Oak Ridge Tokamak experimental power reactor study scoping report

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, M.

    1977-03-01

    This report presents the scoping studies performed as the initial part of the program to produce a conceptual design for a Tokamak Experimental Power Reactor (EPR). The EPR as considered in this study is to employ all systems necessary for significant electric power production at continuous high duty cycle operation; it is presently scheduled to be the final technological step before a Demonstration Reactor Plant (Demo). The scoping study tasks begin with an exploration and identification of principal problem areas and then concentrate on consideration and evaluation of alternate design choices for each of the following major systems: Plasma Engineering and Physics, Nuclear, Electromagnetics, Neutral Beam Injection, and Tritium Handling. In addition, consideration has been given to the integration of these systems and requirements arising out of their incorporation into an EPR. One intent of this study is to document the paths explored in search of the appropriate EPR characteristics. To satisfy this intent, the explorations are presented in chart form outlining possible options in key areas with extensive supporting footnotes. An important result of the scoping study has been the development and definition of an EPR reference design to serve as (1) a common focus for the continuing design study and (2) a guide for associated development programs. In addition, the study has identified research and development requirements essential to facilitate the successful conceptual design, construction, and operation of an EPR.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

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

  5. Overview of light sources powered by tritium

    International Nuclear Information System (INIS)

    Due to their long lifespan and stable intensity, light sources initiated by tritium instead of electricity or batteries are suitable for low level lighting applications. Therefore, tritium-based radioluminescent (RL) light sources are widely used in both military and civil applications. However, traditional tritium lights with the gas tube structure have several shortcomings: (1) the phosphors are opaque; (2) the glass tube is fragile and easily broken; and (3) the beta kinetic energy is attenuated due to the sorption by the gas; etc. As a result, further application of the tritium lights is limited. In this paper, the lighting mechanism and radiation safety of tritium-based RL light sources are briefly reviewed. Besides, the history and prospects of the development of tritium-based RL light source are discussed. Due to their long lifespan and stable intensity, light sources initiated by tritium instead of electricity or batteries are suitable for low level lighting applications. Therefore, tritium- based radioluminescent (RL) light sources are widely used in both military and civil applications. However, traditional tritium lights with the gas tube structure have several short- comings: (1) the phosphors are opaque; (2) the glass tube is fragile and easily broken; and (3) the beta kinetic energy is attenuated due to the sorption by the gas; etc. As a result, further application of the tritium lights is limited. In this paper, the lighting mechanism and radiation safety of tritium-based RL, light sources are briefly reviewed. Besides, the history and prospects of the development of tritium-based RL light source are discussed. (authors)

  6. Development of Environmentally-Assisted Fatigue Monitoring System for Advanced Power Reactors (APR1400)

    Energy Technology Data Exchange (ETDEWEB)

    Park, June Soo; Kim, Yeon Jeong; Kang, Sun Yeh; Yoon, Ki Seok; Choi, Taek Sang [KEPCO-E and C, Daejeon (Korea, Republic of)

    2013-10-15

    This paper introduces an EAF monitoring system developed for Shin-Kori Nuclear Power Plant (NPP), Units 3 and 4 which are the first two reactors of the APR1400 model. The EAF monitoring system has been developed for Shin-Kori NPP, Units 3 and 4, and is ready for an application for the plant lifetime. It is expected that the plant fatigue management can be effectively fulfilled, and the structural integrity of the critical components assured by an implementation of the fatigue monitoring system from the beginning of the lifetime. When fatigue analyses including the effects of the Light-Water Reactor (LWR) environment are applicable, plant designers address the environmentally-assisted fatigue (EAF) for Class 1 reactor pressure boundary components. The environment factor (F{sub en}) method has been endorsed by the U. S. Nuclear Regulatory Commission for evaluating fatigue analyses to address the environmental effects, and this method considers four major variables in addition to the traditional air-fatigue analyses: Material temperature, dissolved oxygen content of coolant, sulfur (S) content of material, and strain rate at the material points of interest. APR1400 nuclear power plants are designed to the requirements of the enhanced plant safety, availability and performance criteria for a 60 year design life. To better manage the material degradation and structural integrity of the pressure boundary components, a fatigue monitoring system has been developed for APR1400 NPPs, which is capable to monitor the EAF damage during the plant lifetime.

  7. Computational Neutronics Methods and Transmutation Performance Analyses for Light Water Reactors

    International Nuclear Information System (INIS)

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One obvious path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCm, PuNpAm, PuNp, and Pu

  8. Review of the status of low power research reactors and considerations for its development

    Energy Technology Data Exchange (ETDEWEB)

    Lim, In Cheol; Wu, Sang Ik; Lee, Byung Chul; Ha, Jae Joo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    At present, 232 research reactors in the world are in operation and two thirds of them have a power less than 1 MW. Many countries have used research reactors as the tools for educating and training students or engineers and for scientific service such as neutron activation analysis. As the introduction of a research reactor is considered a stepping stone for a nuclear power development program, many newcomers are considering having a low power research reactor. The IAEA has continued to provide forums for the exchange of information and experiences regarding low power research reactors. Considering these, the Agency is recently working on the preparation of a guide for the preparation of technical specification possibly for a member state to use when wanting to purchase a low power research reactor. In addition, ANS has stated that special consideration should be given to the continued national support to maintain and expand research and test reactor programs and to the efforts in identifying and addressing the future needs by working toward the development and deployment of next generation nuclear research and training facilities. Thus, more interest will be given to low power research reactors and its role as a facility for education and training. Considering these, the status of low power research reactors was reviewed, and some aspects to be considered in developing a low power research reactor were studied.

  9. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  10. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  11. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  12. Compensation by RGMS for misreading reactor power in case of D2O dilution

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sang Hoon; Park, Jae Yoon; Choi, Young San; Kim, Young Ki [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    In a research reactor Neutron Measurement System (NMS) which uses wide range fission chamber as neutron detector is applied to measure the reactor power. This system has rapid response to power and stable accuracy for wide range. But this has some concerns of relative measured values depending on the installed location of neutron detector and also may cause the loss of accuracy when dilution of heavy water in the D2O tank happens. The NMS is not only used for reactor control and but also used for reactor protection system. Accordingly faulted reactor power with high deviation for second case may lead unexpected increase of the reactor power. In order to prevent this occurrence, Reactor Gamma Measurement System (RGMS) is necessarily applied. Herein the structure, measuring method and application of RGMS will be introduced.

  13. A preliminary investigation of the Topaz II reactor as a lunar surface power supply

    Energy Technology Data Exchange (ETDEWEB)

    Polansky, G.F. [Sandia National Labs., Albuquerque, NM (United States); Houts, M.G. [Los Alamos National Lab., NM (United States)

    1995-12-31

    Reactor power supplies offer many attractive characteristics for lunar surface applications. The Topaz II reactor resulted from an extensive development program in the former Soviet Union. Flight quality reactor units remain from this program and are currently under evaluation in the United States. This paper examines the potential for applying the Topaz II, originally developed to provide spacecraft power, as a lunar surface power supply.

  14. COMSORS: A light water reactor chemical core catcher

    International Nuclear Information System (INIS)

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B2O3) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

  15. Injecting light of high-power LEDs into thin light guides

    NARCIS (Netherlands)

    Cornelissen, H.J.; Ho, C.; Ma, H.; Krijn, M.P.C.M.; Van Sprang, H.A.

    2010-01-01

    A new method using a thin-film multilayer filter is described to couple light from high-power LEDs into a thin light guide such as an LCD backlight. Light emitted below the critical angle is reflected back to the LED and recycled. Largeangle emitted light passes the filter and is transported by tota

  16. Injecting Light of High-Power LEDs into Thin Light Guides

    NARCIS (Netherlands)

    Cornelissen, H.J.; Krijn, M.P.C.; Ma, H.; Van Sprang, H.A.

    2010-01-01

    A new method using a thin-film multilayer filter is described to couple light from high-power LEDs into a thin light guide such as an LCD backlight. Light emitted below the critical angle is reflected back to the LED and recycled. Large-angle emitted light passes the filter and is transported by tot

  17. General Atomic Company fusion experimental power reactor conceptual design

    International Nuclear Information System (INIS)

    The results of a two-year, conceptual design study of a fusion experimental power reactor (EPR) are presented. For this study, the primary objectives of the EPR are to obtain plasma ignition conditions and produce net electrical power. The design features a Doublet plasma configuration with a major radius of 4.5 meters. The average plasma beta is 10 percent which yields a thermonuclear power level of 410 MW during a 105 second burn period. With a duty factor of 0.84, the gross electrical output is 124 MW(e) while the net output is 37 MW(e). The design features a 25 cm thick, helium cooled, modular, stainless steel blanket with a 1 cm thick, thermal radiation-cooled silicon carbide first wall. Sufficient shielding is provided to permit contact maintenance outside the shield envelop within 24 hours after shutdown. An overall facility concept was developed, including a superheated steam cycle power conversion system. Preliminary cost estimates and construction schedules were also developed

  18. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  19. Energy analysis and carbon dioxide emission of Tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Energy gain and carbon dioxide (CO2) emission of tokamak fusion power reactors are evaluated in this study compared with other reactor types, structural materials, and other Japanese energy sources currently in use. The reactors treated in this study are (1) a conventional physics performance international thermonuclear experimental reactor (ITER), like a reactor based upon the ITER engineering design activity (ITER-EDA), (2) a RS (reversed shear) reactor using the reversed shear safety-factor/plasma current profile, and (3) a ST (spherical torus) reactor based upon the final version of the advanced reactor innovative engineering study ST (ARIES-ST). The input energy and CO2 emission from these reactors are calculated by multiplying the weight or cost of the fusion reactor components by the energy intensity and/or with the CO2 intensity data, which are updated as often as possible. The ITER cost estimation is estimated based on the component unit costs. The following results were obtained: (1) The RS and the ST reactor can double the energy gain and reduce CO2 emission by one-half compared with the ITER-like reactor. (2) Silicon carbide (SiC) used as the structural material of inner vessel components is best for energy gain and CO2 emission reduction. (3) The ITER-like reactor is slightly superior to a photovoltaic (PV) with regard to CO2 emission. (4) The energy gain and CO2 emission intensity of the RS reactor and the ST reactor are as excellent as those of a fission reactor and a hydro-powered generator. These results indicate that a tokamak fusion power reactor can be one of the most effective power-generating technologies both in high-energy payback gains and reduction of CO2

  20. Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  1. Refurbishment Status of Light Ⅱ-A Pulsed Power Generator

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>The Light Ⅱ-A pulsed power generator, formerly used to pump KrF laser, was updated around the year 2000 from the original Light Ⅱ, a pulsed power generator built in 1980s at CIAE. This machine was

  2. Study of Pu consumption in Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology

  3. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  4. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    International Nuclear Information System (INIS)

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

  5. Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor

    Directory of Open Access Journals (Sweden)

    Fan Zhang

    2014-01-01

    Full Text Available SCWR (Supercritical Water Reactor is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was used to calculate the thermodynamic properties of water and steam. The ISO-5167-4: 2003 standard was incorporated in the code as the basis of orifice plate to compute the flow rate. New heat balance model and uncertainty estimate have also been included in the code. In order to validate H-Power, an assessment was carried out by using data published by US and Qinshan Phase II. The results showed that H-Power was able to estimate the thermal power of SCWR.

  6. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    International Nuclear Information System (INIS)

    The technology of breeding 233U from 232Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program

  7. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    Energy Technology Data Exchange (ETDEWEB)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  8. Questions to the reactors power upgrade of the Nuclear Power Plant of Laguna Verde

    International Nuclear Information System (INIS)

    The two reactors of the Nuclear Power Plant of Laguna Verde (NPP-L V) were subjected to power upgrade labors with the purpose of achieving 20% upgrade on the original power; these labors concluded in August 24, 2010 for the Reactor 1 and in January 16, 2011 for the Reactor 2, however in January of 2014, the NNP-L V has not received by part of the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) the new Operation License to be able to work with the new power, because it does not fulfill all the necessary requirements of safety. In this work is presented and analyzed the information obtained in this respect, with data provided by the Instituto Federal de Acceso a la Informacion Publica y Proteccion de Datos (IFAI) and the Comision Federal de Electricidad (CFE) in Mexico, as well as the opinion of some workers of the NPP-L V. The Governing Board of the CFE announcement that will give special continuation to the behavior on the operation and reliability of the NPP-L V, because the frequency of not announced interruptions was increased 7 times more in the last three years. (Author)

  9. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    International Nuclear Information System (INIS)

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

  10. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2014-06-15

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

  11. Method of locating a leaking fuel element in a fast breeder power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, John R. (Downers Grove, IL); Fryer, Richard M. (Idaho Falls, ID)

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  12. Method of locating a leaking fuel element in a fast breeder power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Honekamp, J.R.; Fryer, R.M.

    1978-03-21

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of /sup 134/Xe to /sup 133/Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  13. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    Science.gov (United States)

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  14. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    Science.gov (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-10-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  15. Power conversion systems based on Brayton cycles for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linares, J.I., E-mail: linares@upcomillas.es [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain); Herranz, L.E. [Unit of Nuclear Safety Research. CIEMAT, Madrid (Spain); Moratilla, B.Y.; Serrano, I.P. [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain)

    2011-10-15

    This paper investigates Brayton power cycles for fusion reactors. Two working fluids have been explored: helium in classical configurations and CO{sub 2} in recompression layouts (Feher cycle). Typical recuperator arrangements in both cycles have been strongly constrained by low temperature of some of the energy thermal sources from the reactor. This limitation has been overcome in two ways: with a combined architecture and with dual cycles. Combined architecture couples the Brayton cycle with a Rankine one capable of taking advantage of the thermal energy content of the working fluid after exiting the turbine stage (iso-butane and steam fitted best the conditions of the He and CO{sub 2} cycles, respectively). Dual cycles set a specific Rankine cycle to exploit the lowest quality thermal energy source, allowing usual recuperator arrangements in the Brayton cycle. The results of the analyses indicate that dual cycles could reach thermal efficiencies around 42.8% when using helium, whereas thermal performance might be even better (46.7%), if a combined CO{sub 2}-H{sub 2}O cycle was set.

  16. Reactivity control rod for controlling reactor power distribution

    International Nuclear Information System (INIS)

    Since a cladding tube is situated at the outer side, it undergoes neutron irradiation in a reactor core and also undergoes compression force due to high pressure of reactor coolants to cause a creep phenomenon, and the diameter is reduced as it is used. Then, neutron absorbing rods as reactivity control rods for controlling the power distribution are constituted with a cladding tube, a spacer tube disposed at the central portion of the cladding tube and a borosilicate glass tube disposed between the cladding tube and the spacer tube. The gap between the borosilicate glass tube and the spacer tube is gradually changed so that the inner diameter of the borosilicate glass is increased as it comes closer to the lower end plug. The time of contact between the cladding tube and the spacer tube in the inside is delayed by the constitution of the borosilicate glass tube disposed in the cladding tube of the neutron absorbing rod as the reactivity control rod thereby capable of extending the integral working life time with no rupture of the cladding tube. (N.H.)

  17. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Science.gov (United States)

    2012-10-12

    ... COMMISSION Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does... Section 19.3, ``Regulatory Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light......

  18. Research reactors

    International Nuclear Information System (INIS)

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

  19. Radioactivity effects of Pb-17Li in fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rocco, P. (Commission of the European Communities, Joint Research Centre, Ispra (Italy)); Zucchetti, M. (Dipt. di Energetica, Politecnico Turin (Italy))

    1991-12-01

    Research on the eutectic Pb-17Li is part of the blanket studies carried out in Europe for fusion power reactors. The use of this breeder makes easier some safety problems as compared to the case of lithium as a consequence of the lower chemical reactivity of Pb-17Li. On the other hand, it increases the radioactivity problems due to the neutron activation of lead and impurities. This paper presents both short-term (accidents) and long-term (waste disposal and recycling) aspects of the Pb-17Li activation products. They include the production, mobilization, release and environmental impact. Concerning accidents, a particular attention is given to Po-210 and Hg-203. Questions related to waste management are also revised. The most attractive solution seems that of recycling the spent Pb-17Li. This will be possible about 20 y after removal from service. As an alternative to recycling, the breeder disposal as radioactive waste is discussed. (orig.).

  20. Non-Power Reactor Operator Licensing Examiner Standards

    International Nuclear Information System (INIS)

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