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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 Background Carolina Power & Light Company (the licensee), now doing business as Progress Energy Carolinas... significant effect on the quality of the human environment (75 FR 3942, dated January 25, 2010)....

  4. 75 FR 77919 - Carolina Power & Light Company Shearon Harris Nuclear Power Plant, Unit 1; Environmental...

    Science.gov (United States)

    2010-12-14

    ... License No. NPF-63, issued to Carolina Power & Light Company (the licensee), now doing business as... part 73 as discussed in a Federal Register notice dated March 27, 2009 (74 FR 13926). There will be no... COMMISSION Carolina Power & Light Company Shearon Harris Nuclear Power Plant, Unit 1;...

  5. 77 FR 13156 - Carolina Power & Light Company; Shearon Harris Nuclear Power Plant, Unit 1; Exemption

    Science.gov (United States)

    2012-03-05

    ... COMMISSION Carolina Power & Light Company; Shearon Harris Nuclear Power Plant, Unit 1; Exemption 1.0... Harris Nuclear Power Plant (HNP), Unit 1. The license provides, among other things, that the facility is...) 50.46, ``Acceptance criteria for emergency core cooling systems for light- water nuclear...

  6. 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

  7. Quality management initiatives at Carolina Power and Light Company

    International Nuclear Information System (INIS)

    At Carolina Power and Light Company nuclear quality assurance functions are performed including quality control of site work processes, surveillance of site activities, corporate auditing and quality assurance engineering. In addition, an independent section performs technical evaluations, field observation activities, and safety system assessments. In 1989, the Company's management chartered a team of experienced nuclear managers to help identify opportunities and specific means to improve the Company's self-assessment processes. The team also visited or surveyed approximately 20 other utilities, INPO, NRC's Regional and Central headquarters in order to identify the best practices in the U.S. nuclear industry to self-assessment. The major conclusions drawn by the Company from this review activity are presented. As a result, a number of specific changes have been implemented at the Company, both organizationally and functionally, e.g. the Corporate Quality Assurance audit programme was refocused to introduce performance-based techniques and philosophy. Some of the changes are briefly described. (Z.S.)

  8. 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

  9. 76 FR 53970 - Carolina Power & Light; Brunswick Steam Electric Plant, Units 1 and 2; Independent Spent Fuel...

    Science.gov (United States)

    2011-08-30

    ... accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process requires... Licenses Nos. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2, including the BSEP Independent Spent Fuel Storage Installation, currently held by Carolina Power & Light Company,...

  10. Decommissioning of light-water reactor nuclear power plants

    International Nuclear Information System (INIS)

    This study deals with the technical and economic questions posed by the decommissioning of light-water reactor nuclear power plants of the 900-1300 MW class, account being taken of the distinctions between boiling- and pressurized-water reactors. Possible decommissioning alternatives and the disposal or confinement of activity are discussed. It emerges from the discussion that decommissioning, and even total dismantlement of these nuclear power plants is in principle feasible. The activity inventory, one year after shutdown, is calculated to be about 3 X 107 Ci for the BWR and 4 X 106 Ci for the PWR; 40 years after shutdown these figures are reduced to 2 X 106 and 4 X 105 Ci, respectively. The decommissioning costs to be expected are also estimated. This estimate serves as the basis for an economic comparison by the present worth method. The economic comparison shows that total dismantlement after a cooling time of one year is more than four times as expensive as interim confinement followed by total dismantlement waiting period of 40 years. The present worths for immediate total dismantlement are estimated at DM 200 million for the BWR and DM 170 million for the PWR; for the other alternative, they are put at DM 45 million for the BWR and DM 42 million for the PWR. A still open question is posed by the final storage of the large quantities of bulky radioactive waste arising in partial or total dismantlement. Since no decision on the storage method has yet been taken, disposal in casks is stipulated as a boundary condition in the estimation of the costs, although this is an unrealistic assumption. It is to be presumed that the costs of disposal can be reduced given appropriate final storage. (Auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Burns, T J; Cheverton, R D; Flanagan, G F; White, J D; Ball, D G; Lamonica, L B; Olson, R

    1986-05-01

    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.

  12. 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

  13. 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...

  14. Generic environmental impact statement on handling and storage of spent light water power reactor fuel. Appendices

    International Nuclear Information System (INIS)

    Detailed appendices are included with the following titles: light water reactor fuel cycle, present practice, model 1000MW(e) coal-fired power plant, increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data (1976-2000), characteristics of nuclear fuel, and ''away-from-reactor'' storage concept

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

    Science.gov (United States)

    2013-10-25

    ... 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 and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    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.

  18. Siting of light-water reactor power plants in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    The nuclear power plant site requirements formulated for environment protection in Germany allow nuclear power plants to be built at any site provided these requirements are duly taken into account in preparing and monitoring the site and in the design of the proposed power plant. After a brief discussion of light water reactor power plant sites, prevailing practice in site planning, site selection criteria, licensing procedure and used criteria, rules and guidelines, this paper reports on some considerations taken into account by the expert advisers and by the licensing authorities and future site planning. (orig.)

  19. Sensitivity of gamma thermometers for local power monitoring in light water reactors

    International Nuclear Information System (INIS)

    Radcal Gamma Thermometers for local power monitoring show certain advantages compared to neutron sensitive detectors that make them attractive for use in light water reactors. They are based on measurements of the total heat rate in the detector due to neutron-and gamma ray absorption. To correlate gamma detector readings to fuel assembly power under actual reactor conditions the gamma detector response function must be known. A method was developed, based on validated computer codes and data libraries, to determine the gamma thermometer sensitivity as a function of the most important reactor parameters such as burn-up, moderator density etc. The gamma thermometer sensitivity determined by theory could be compared by corresponding experimentally derived values for the Swedish Ringhals-2 PWR. The sensitivity mainly depends on the burn-up of the instrumented assembly, which could be explained by theory and verified by the experimentally obtained results

  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

    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

  2. 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

  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. Trends in development of in-core monitoring of light water cooled reactor power distribution

    International Nuclear Information System (INIS)

    Trends and achievements in the development of in-core monitoring of light water cooled reactor power distribution are described. Application of inertialess detectors in control systems of the 1300 MW PWR reactors operating in the load-following regime, various options of the systems, the capabilities and prospects of further improvement are considered. Problems accompanying their application are analysed. Wire and fall-type activation detectors were used in the PWR and BWR reactors at the initial stages of in-core monitoring system development. Later systems with stationary β-emission neutron detectors, with rhodium or vanadium emitters and compton emission neutron detectors with emitters of cobalt or hafnium gained wide application. Gamma thermometer which require no calibration during the operation appear to be new detectors for continuous control in PWRs. However they are intertice and their application does not rid of the necessity to have sensors with minor signal delay in the control system and does not allow one to refuse periodic changes of axial power by mobile detectors, providing for high accuracy and detalization of distributions

  5. Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

    International Nuclear Information System (INIS)

    The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors)

  6. 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.

  7. 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

  8. New approach for control rod position indication system for light water power reactor

    International Nuclear Information System (INIS)

    Control rod position indication system is an important system in a nuclear power plant to monitor and display control rod position in all regimes of reactor operation. A new approach to design a control rod position indication system for sensing absolute position of control rod in Light Water Power Reactor has been undertaken. The proposed system employs an inductive type, hybrid measurement strategy providing both analog position as well as digital zone indication with built-in temperature compensation. The new design approach meets single failure criterion through redundancy in design without sacrificing measurement resolution. It also provides diversity in measurement technique by indirect position sensing based on analysis of drive coil current signature. Prototype development and qualification at room temperature of the control rod position indication system (CRPIS) has been demonstrated. The article presents the design philosophy of control rod position indication system, the new measurement strategy for sensing absolute position of control rod, position estimation algorithm for both direct and indirect sensing and a brief account associated processing electronics. (author)

  9. 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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 and 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

  13. 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.

  14. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work

  15. 76 FR 54261 - Carolina Power & Light; H.B. Robinson Steam Electric Plant, Unit No. 2; HBRSEP Independent Spent...

    Science.gov (United States)

    2011-08-31

    ... accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process requires... pursuant to 10 CFR 72.50 for Robinson ISFSI, along with Brunswick Steam and Electric Plant (BSEP), Units 1 and 2, BSEP Independent Spent Fuel Storage Installation, Shearon Harris Nuclear Power Plant, Unit...

  16. 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.

  17. 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.)

  18. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    International Nuclear Information System (INIS)

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions (235U and 238U fissions and 238U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D2O in moderator (FD2O) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and FD2O that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  19. Calculation of releases of radioactive materials in gaseous and liquid effluents from light-water-cooled power reactors

    International Nuclear Information System (INIS)

    This regulatory guide references two NUREG reports that provide acceptable methods for calculating annual average expected releases of radioactive material in liquid and gaseous effluents from light-water-cooled nuclear power reactors. The procedures and models provided in the referenced NUREG reports will be subject to continuing review by the NRC staff with the aim of employing the best available experimental data and calculational models in order to achieve increased accuracy and realism. As a result of such reviews, it is expected that alternative acceptable methods for calculation will be made available to applicants and that calculational procedures found to be unnecessary will be eliminated. The guide supersedes portions of Regulatory Guide 1.42, Revision 1, Interim Licensing Policy on as Low as Practicable for Gaseous Radioiodine Releases from Light-Water-Cooled Nuclear Power Reactors, which has been withdrawn

  20. Advanced light water reactor program at ABB-Combustion Engineering Nuclear Power

    International Nuclear Information System (INIS)

    To meet the needs of Electric Utilities ordering nuclear power plants in the 1990s, ABB-Combustion Engineering is developing two designs which will meet EPRI consensus requirements and new licensing issues. The System 80 Plus design is an evolutionary pressurized water reactor plant modelled after the successful System 80 design in operation in Palo Verde and under construction in Korea. System Plus is currently under review by the US Nuclear Regulatory Commission with final design approval expected in 1991 and design certification in 1992. The Safe Integral Reactor (SIR) plant is a smaller facility with passive safety features and modular construction intended for design certification in the late 1990s. (author)

  1. 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

  2. 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

  3. 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

  4. Light water type reactor

    International Nuclear Information System (INIS)

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

  5. 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

  6. 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

  7. Reactor power measuring device

    International Nuclear Information System (INIS)

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

  8. Design study on a very long life light water power reactor core

    International Nuclear Information System (INIS)

    Aiming of effective utilization of plutonium, an idea of a very high burnup light water reactor (VHBLWR) core was proposed, and new concept of the fuel cycle, in which the breeding of TRU nuclides will be restrained (TRU Enclosing Fuel Cycle), was discussed. The VHBLWR core consists of highly enriched plutonium mixed-oxide (MOX) fuel rods and 238UO2 rods. A burnup of more than 100 GWd/t can be expected with small excess reactivity and negative void coefficient throughout the core life, and it is clarified that TRU nuclides can play a role of the burnable poison in the VHBLWR. It is also clarified that the TRU Enclosing Fuel Cycle is feasible, by burning TRUs with MOX fuel, repeatedly in the VHBLWRs. Studies on performances of fuel and materials with long-term irradiation, and on items related to fuel-reprocessing, fuel-transportation, MOX fuel-fabrication, etc, were carried out. (author)

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

    International Nuclear Information System (INIS)

    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)

  10. 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).

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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)

  17. 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)

  18. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention efficiently calibrates a fixed type gamma ray thermometer of a reactor power measuring device of a BWR type reactor. Namely, the device of the present invention calculates peripheral fuel rod power distribution by calibrating the reactor power distribution by heat generation amount, the reactor power distribution being obtained by a calculation based on a reactor model for converting the signals of a plurality of the gamma ray thermometers in the reactor core based on a conversion formula. In this case, the conversion formula is a relational formula between the power of a thermocouple of the gamma ray thermometer, gamma ray heat generation amount, thermocouple zero power sensitivity relative to a temperature coefficient. A conversion efficient calculation means makes a calibration heater to generate heat at a predetermined power, and the thermocouple zero power sensitivity and the temperature coefficient are obtained based on the output of the gamma ray thermometer in this case. The calibration means updates to conversion type thermocouple zero power sensitivity and temperature coefficient. A calibration execution means executes the operations described above successively, and when the thermocouple zero power sensitivity and the temperature coefficient are out of an allowable range, the means informs it and eliminates the corresponding gamma ray thermometer from the measuring meters. (I.S.)

  19. 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.

  20. 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

  1. Reactor power control device

    International Nuclear Information System (INIS)

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

  2. 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.)

  3. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention concerns a method of controlling reactor power to shift it into a partial power operation upon occurrence of recycling pump tripping or loss of generator load. Operation state of a reactor is classified into a plurality of operation states based on values of the reactor core flow rate and the reactor power. Different insertion patterns for selected control rods are determined on every classified operation states. Then, an insertion pattern corresponding to the operation state upon occurrence of recycling pump tripping or loss of power generator load is carried out to shift into partial power operation. The operation is shifted to a load operation solely in the station while avoiding risks such as TPM scram. Then neutron fluxes are suppressed upon transient to increase margin of fuel integrity. Selected control rod pattern of the optimum reactivity is set to each of operation regions, thereby enabling to conduct flexible countermeasure so as to attain optimum operationability. (N.H.)

  4. 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.

  5. Present state of new technologies of nuclear power generation, and technological development of fast-breeder reactor and next-generation light water reactor

    International Nuclear Information System (INIS)

    This paper introduces the present state of development of FBR in Japan and international cooperation, the development of HP-ABWR and HP-APWR as the next-generation light water reactors, and SMR development in the United States. As for FBR, the following situations are described: (1) history of development in Japan in the past, (2) history of change due to the readjustment of development plan caused by the accident of Fukushima Daiichi Nuclear Power Station, in which shift to FaCT phase 2 was suspended, and the approach to the establishment of safety standards for sodium-cooled FBR and its international standardization was adopted, and (3) future challenges. As for the Japan - France fast-breeder reactor development cooperation, the conclusion of the Japan - France inter-government agency agreement, and Japan's cooperation plan and system are described. Next, as for HP-ABWR and HP-APWR, the development goal and concept of each plant, and the element technologies required for the success are described. On the other hand, the small reactor development in the United States started with the aim of the securement of domestic technology base, contribution to reduction in carbon dioxide emissions, and its export to new entry countries for nuclear energy. This project aimed the practical use of SMR, and started 'financial support program for small reactors' to allocate about 452 million dollars to maximum two units of SMRs in the next five years. This project is outlined. (A.O.)

  6. 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)

  7. 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.

  8. 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

  9. INB 40 OSIRIS and ISIS reactors: complementary safety evaluation in the light of the Fukushima Daiichi nuclear power station accident

    International Nuclear Information System (INIS)

    This report proposes a complementary safety evaluation of the Osiris and Isis reactors (INB 40) in Saclay, one of the French basic nuclear installations (BNI, in French INB) in the light of the Fukushima accident. This evaluation takes the following risks into account: risks of flooding, earthquake, loss of power supply and loss of cooling, in addition to operational management of accident situations. It presents some characteristics of the installation (brief description, buildings, effluents, control-command, radioprotection), identifies the risks of cliff effect and the main structures and equipment, evaluates the seismic risk (installation sizing, installation conformity, margin evaluation), evaluates the flooding risk (installation sizing, installation conformity, margin evaluation), briefly examines other extreme natural phenomena (extreme events, combination of events, earthquake or flooding with a higher level than that for which the installation is designed, measures to prevent a cliff effect). It analyzes the risk of a loss of power supply and of cooling (loss of external and internal electric sources, loss of the ultimate cooling system). It analyzes the management of severe accidents: crisis management organization, training and exercises, available intervention means, robustness of available means, measures for the protection of the confinement integrity after fuel damage. It discusses the conditions of the use of subcontractors

  10. Results of research and development for nuclear power plants with WWER-1000 type light water reactors

    International Nuclear Information System (INIS)

    The conference met in three sessions: 1. Project designing and construction of nuclear power plants; 2. Materials, technologies and applied mechanics; 3. Physics, thermal physics and control. The proceedings contains 82 papers of which only two have not been inputted in INIS. The final resolutions of session 1 related to the reduction of capital costs for newly built units, processing of project documentation, the introduction of step motors manufactured in Czechoslovakia, in-service diagnostics of nuclear power plants, etc. The final recommendations of session 2 dealt with the centralization of the management of research into the reliability, safety and residual life of nuclear installations, with radiation stability of weld metals, repairs of nuclear power plants by patch welding, with welding in nuclear power plants and stress calculations using mathematical methods. Session 3 centred on questions of the safety, reliability and economy of nuclear power plant operation. It was recommended to make a comparison of the results of theoretical calculations with experiments, to concentrate on the automation of measurement, to extend international division of labour and cooperation of CMEA countries, to extend publishing activities in the field of thermal physics, etc. General recommendations were related to the conception of the construction of nuclear power plants in Czechoslovakia, the implementation of original scientific, research and development work, to the question of personnel for nuclear research, the experimental base of the Czechoslovak nuclear programme and to planning and management of technical development. (E.S.)

  11. Reliability demonstration test for light water reactor type of nuclear power plant

    International Nuclear Information System (INIS)

    When the construction of a nuclear power station is materialized, the question and anxiety about the safety and environmental preservation are presented from the residents in the region, and it is an important policy of the nation to obtain the understanding of people about atomic energy by eliminating such anxiety. The demonstration test of the safety and reliability of machinery and equipments for LWR power stations is planned mainly by the Atomic Energy Technology Test Center. The demonstration test of the reliability of fuel assemblies for BWRs and PWRs, the thermal load test of fuel assemblies, the demonstration test of the reliability of valves, recirculating pumps for BWRs, primary coolant pumps for PWRs, the heat-affected zones of welded parts in pipings, and in-service inspection are outlined. In Japan where earthquakes frequently occur, sufficient consideration is given to the aseismatic design of nuclear power generation facilities, and the construction of a large, high performance oscillating platform is in progress in Tadotsu, Kagawa Prefecture. The demonstration test of aseismatic reliability will be started in 1981. The demonstration test of the reliability of steam generators for PWR power stations has been forwarded since 1975 by the Association on Thermal Engines for Power Generation. Additional demonstration test items are requested. (Kako, I.)

  12. Education and training activities at North Carolina State University's PULSTAR reactor

    International Nuclear Information System (INIS)

    Research reactor utilization has been an integral part of the North Carolina State University's (NCSU's) nuclear engineering program since its inception. The undergraduate curriculum has a strong teaching laboratory component. Graduate classes use the reactor for selected demonstrations, experiments, and projects. The reactor is also used for commercial power reactor operator training programs, neutron radiography, neutron activation analysis (NAA), and sample and tracer activation for industrial short courses and services as part of the university's land grant mission. The PULSTAR reactor is a 1-MW pool-type reactor that uses 4% enriched UO2 pellet fuel in Zircaloy II cladding. Standard irradiation facilities include wet exposure ports, a graphite thermal column, and a pneumatic transfer system. In the near term, general facility upgrades include the installation of signal isolation and computer data acquisition and display functions to improve the teaching and research interface with the reactor. In the longer term, the authors foresee studies of new core designs and the development of beam experiment design tools. These would be used to study modifications that may be desired at the end of the current core life and to undertake the development of new research instruments

  13. 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

  14. Two-dimensional calculation of neutron flux and power distribution in the fuel assembly of light water reactor

    International Nuclear Information System (INIS)

    The accurate determination of the neutron flux (power) distribution in the fuel assemblies is one of the important problems for the nuclear design of light water reactor (LMR). Due to the strong heterogeneity of the cell structure in the assembly of LWR, for the neutron flux calculation in the fuel assembly the diffusion theory will introduce severe errors and we have to solve the neutron transport equation by using some more sophisticated and efficient methods (for example MC or Sn method etc.) but they are very time consuming. Recently some more efficient approach of the collision probability method, the transmission probability method (TPM), has been developed. In this paper, the application of TPM to the calculation of neutron flux distribution in the assembly is described. A simplified and accurate mathematical model is proposed, in which all expansion coefficients of angular flux are determined by the incoming and outgoing currents at region surfaces, and they are renewed and consecutively approach to the exact values following each iteration. Thus, in comparison with usual TPM methods in which the expansion coefficients are determined by weight-residual or functional minimized method, the numerical calculations are much simplified and computing time is reduced substantially

  15. 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

  16. 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

  17. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

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

  18. Operating US power reactors

    International Nuclear Information System (INIS)

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of Dec. 31, 1986, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (October, November, and December 1986) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. They are defined as follows: In addition to the tabular data, this article discusses significant occurrences and developments that affected licensed US power reactors during this reporting period. It includes, but is not limited to, changes in operating status, regulatory actions and decisions, and legal actions involving the status of power reactors. We do not have space here for routine problems of operation and maintenance, but such information is available at the Nuclear Regulatory Commission (NRC) Public Document Room, 1717 H Street, NW, Washington, DC 20555. Some significant operating events are summarized elsewhere in this section in the article ''Selected Safety-Related Events,'' and a report on activities relating to facilities still in the construction process is given in the article ''Status of Power-Reactor Projects Undergoing Licensing Review'' in the last section of each issue of this journal. The reader's attention is also called to the regular feature ''General Administrative Activities,'' which deals with more general aspects of regulatory and legal matters that are not covered elsewhere in the journal

  19. 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.

  20. 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.

  1. 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

  2. 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)

  3. 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

  4. 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

  5. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention comprises a γ-thermometer disposed in a BWR type reactor, a first amplifier for amplifying the output thereof, a fission ionization chamber disposed in the reactor separately from the γ-thermometer, a second amplifier for amplifying the output thereof, a differential circuit for differentiating the output signal of the second amplifier and a first adding circuit for adding an output signal of the differential circuit and an output signal of the first amplifier. Alternatively, a γ-ray self-powered neutron detector may be disposed instead of the fission ionization chamber. A second adding circuit is also disposed for adding the output signals of plurality of differentiation circuits and inputting the result to the first adding circuit. A sensitivity controller is disposed upstream of the first adder for controlling the sensitivity of the fission ionization chamber. Then, even if time delay should be caused in the γ-thermometer, output signals with good time response characteristic can be obtained by using signals of LPRM or SPND, and currently changing output of the reactor can be measured accurately to provide an effect on the improvement of the safety and operation controllability of the reactor. (N.H.)

  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. 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.)

  8. High performance light water reactor

    International Nuclear Information System (INIS)

    The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

  9. 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.

  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. 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

  13. 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.

  14. Analysis on damages of piping and the others in the light water reactor type nuclear power stations due to thermal fatigue

    International Nuclear Information System (INIS)

    As thermal fatigue is an important item in the nuclear energy field, it is not only limited to nuclear energy, but also conducted by a lot of efforts in the other industries. In the nuclear energy field, as receiving thermal change at every start stop of operation in light water reactor type nuclear power station such as boiling water type and pressurized water type, thermal fatigue is anxious to form at its apparatus and piping, by which damages directly relate to their soundness. Therefore, in their design standards to protect such damage in every creation, some regulations on the thermal fatigue have been prepared. In Japan as well, 'technical standard on structures and so on relating to nuclear energy facilities for power generation' according to the Business Rule on Electricity was established to show fatigue limit on every material for design of apparatus to protect to excess the limit during its operation. Here was shown types and forming conditions on damage of apparatus and so on in the light water reactor type power stations due to thermal fatigue by collecting, classifying and arranging some actual examples on the damage formed at the power stations in and out of Japan. (G.K.)

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO2, ThO2, ThO2 -- UO2, ZrO2 -- UO2) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO2, ZrO2 -- UO2 fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO2 and ThO2 -- UO2 fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system

  20. 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

  1. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  2. Requirements for light water reactors

    International Nuclear Information System (INIS)

    The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10-6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

  3. Advanced light water reactor plant

    International Nuclear Information System (INIS)

    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)

  4. Automatic reactor power control device

    International Nuclear Information System (INIS)

    Anticipated transient without scram (ATWS) of a BWR type reactor is judged to generate a signal based on a reactor power signal and a scram actuation demand signal. The ATWS signal and a predetermined water level signal to be generated upon occurrence of ATWS are inputted, and an injection water flow rate signal exhibiting injection water flow rate optimum to reactor flooding and power suppression is outputted. In addition, a reactor pressure setting signal is outputted based on injection performance of a high pressure water injection system or a lower pressure water injection system upon occurrence of ATWS. Further, the reactor pressure setting signal is inputted to calculate opening/closing setting pressure of a main steam relief valve and output an opening setting pressure signal and a closure setting pressure signal for the main steam relief valve. As a result, the reactor power and the reactor water level can be automatically controlled even upon occurrence of ATWS due to failure of insertion of all of the control rods thereby enabling to maintain integrity and safety of the reactor, the reactor pressure vessel and the reactor container. (N.H.)

  5. 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

  6. Extended Cooling System for High Power Reactors

    International Nuclear Information System (INIS)

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants (NPPs) and proposed for advanced light water reactors (LWRs). However, it is not clear that currently proposed external reactor vessel cooling (ERVC) could provide sufficient heat removal for higher power reactors. This paper proposes a dual retention strategy to realize fail-proof defense-in-depth in the APR1400 (Advanced Power Reactor 1400 MWe) and the OPR 1000 (Optimized Power Reactor 1000 MWe). The dual retention has the advantage of IVR-ERVC as well as ex-vessel cooling (EVC) strategies. The multilateral, multidisciplinary project calls for national and international cutting-edge technologies to research and produce (R and P) the D2R2 (Duel Retention Demonstration Reactor) equipped with OASIS (Optimized Advanced Safety Injection System) and ROSIS (Reactor Outer Safety Injection System) to cope with design-basis accidents and beyond in a coherent, continual, comprehensive manner. The enterprise aims to develop the design-basis and severe accident engineering solutions. The enterprise aims to develop the design-basis and severe accident engineering solutions. The former embraces ISAIAH (Injection System Annular Interactive Aero Hydrodynamics) and MESIAH (Methodical Evaluation System Interactive Aero Hydrodynamics). The latter comprises GODIVA (Geo metrics of Direct Injection Versatile Arrangement), SONATA (Simulation of Narrow Annular Thermomechanical Arrest or), TOCATA (Termination of Corium Ablation Thermal Attack) and STRADA (Solution to Reactor Advanced Design Alternatives). D2R2 will contribute to enhancement of both safety and economics for an advanced high power particular and nuclear power in general

  7. 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

  8. Safety of light water reactors. Risks of nuclear technology

    International Nuclear Information System (INIS)

    The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

  9. Low-power nuclear reactors

    International Nuclear Information System (INIS)

    A brief development history of low-power nuclear reactors is presented in this paper. Nowadays, some countries have plans to build a series of small nuclear power plants (also floating ones) for use in remote regions. Present constructions of such NPP are presented in this paper. (author)

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

    Science.gov (United States)

    2010-01-01

    ...-ended rupture of the largest pipe in the reactor coolant system. (2) An evaluation model is the... cladding locally were converted to stoichiometric zirconium dioxide. If cladding rupture is calculated to... calculated time of rupture. Cladding thickness before oxidation means the radial distance from inside...

  11. 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.)

  12. Safety of Generation IV reactors in light of engineering findings from the Fukushima Dai-ichi Nuclear Power Station accident

    International Nuclear Information System (INIS)

    Public disclosure of technical information about the safety of Generation IV sodium-cooled fast reactors (SFR) was planned by the AESJ in autumn 2012. SFR was an important key technology to assure long-term energy supply in Japan and an international society prepared for uncertain energy security in the future. After the Fukushima Daiichi severe accident, it was highly important to learn the lessons and technical knowledge. This article described (1) outline of safety design criteria (SDC) under preparation by Generation IV International Forum (GIF), (2) analysis of technical knowledge of the Fukushima Daiichi accident for Generation IV reactor, (3) severe accident management and its effectiveness and (4) safety assuring measures reflecting from lessons of the Fukushima accident. (T. Tanaka)

  13. Management of radioactive wastes at power reactor sites in India

    International Nuclear Information System (INIS)

    Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

  14. Reactor regulating and protection system for a light water reactor

    International Nuclear Information System (INIS)

    Microprocessor based systems are developed for reactor regulation and protection of LWR. A triple modular redundancy approach is followed for the design of this system. This system is functionally partitioned into two sub-systems - Reactor Regulating System (RRS) and Reactor Trip Logic System (RTLS). RRS controls the reactor power as per demand and RTLS generates the reactor trip on abnormal process conditions. This paper describes the details of RRS and RTLS system architecture and fault tolerant and fail-safe features used in the system design. (author)

  15. Cascade ICF power reactor

    International Nuclear Information System (INIS)

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO2. The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor

  16. 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

  17. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    When occurrence of earthquakes is judged in a BWR type reactor, the power is decreased by inserting a portion of control rods, reducing a speed of recycling pumps, stopping recycling pumps, increasing the opening degree of a main steam control valve and opening a main steam relief valve. The reactor scram can be avoided by bypassing neutron flux high signal, settling a filter to neutron flux signals and setting a reactor scram set value by neutron flux signals, for example, to 120%. There is constituted an interlock for performing reactor scram when both of a neutron flux high signal and a signal outputted if a surface heat flux corresponding signal formed by applying calculation to the neutron flux high signal exceeds a set value are valid, to avoid unnecessary reactor scram. As a measuring means, not only an acceleration meter in the power plant, but also acceleration meters at remote places, acceleration meters or displacement meters for various kinds of equipments in the power plant are used, and when signals from them exceed set values, earthquake judgement is conducted. (N.H.)

  18. 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

  19. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

    Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

  20. 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

  1. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  2. Can light water reactors be proliferation resistant?

    International Nuclear Information System (INIS)

    During the last decade several questions were raised concerning the proliferation issues of Light Water Reactors (LWRs) in comparison with other types of power reactors, particularly Gas Cooled Reactors (GCRs) and Heavy Water Reactors (HWRs). These questions were strongly highlighted when the Agreed Framework between the United States and the DPRK was signed in October 1994 and following the formation of KEDO organization to provide two LWRs to DPRK in replacement of all its GCRs in its nuclear program. One might summarize the main questions into three groups, mainly: 1. Can LWRs produce weapon-grade Plutonium (Pu)? 2. Why is the LWR type considered as a better option with regard to non-proliferation compared to other power reactors - particularly GCR and HWR types? 3. How could LWRs be more resistant to proliferation? This paper summarizes the effort to answer these questions. Included tables present numerical parameters for Pu production capability of the three main reactor types (LWRs, GCRs and HWRs) of a 400 MWe power reactor unit, during normal operation, and during abnormal operation to produce weapon grade Pu. Can LWRs produce weapon-grade Pu? It is seen from the available data that weapon-grade Pu could be produced in LWR fuel, as in the fuel of most other power reactor types, by limiting fuel irradiation to two or three months only. However, such production, though possible, is exceptional. In a recent study 5% of LWRs under IAEA safeguards have spent fuel inventory containing limited amount of high-grade Pu. The equilibrium burnup of discharged fuel is in the order of 33,000 MWD/T. However and due to lower enrichment of initial inventory almost half of that burnup is produced. In normal situations the discharged initial inventory has a Pu grade which is less than weapon grade and is unlikely to be used for weapon production. Why LWR the type is considered as a better option for non-proliferation Referring to tables, one can conclude that LWRs make less Pu

  3. 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)

  4. 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

  5. Radiation damage of light water reactor materials

    International Nuclear Information System (INIS)

    Reactor materials and irradiation conditions have changed as reactor types advanced from light water reactors to fast breeder reactors and future nuclear fusion reactors, but the physical and chemical processes of irradiation damage at the final stage of service of the structural materials of light water reactors must be studied more in detail. The research on the secular change of LWR materials under irradiation is regarded as most important. Especially in the problems of the embrittlement of pressure vessels, the irradiation damage of in-core structures, and the irradiation growth and creep of zircaloy, it has become more important to continuously evaluate the secular deterioration and assure the soundness, and to elucidate the process of microstructure chanage of the materials under irradiation for forecasting the life. The structure and the materials of light water reactors are explained. The neutron irradiation embrittlement of reactor pressure vessels must be forecast more accurately as the years of operation elapse. The factors affecting the irradiation embrittlement are described. The physical model and the microstructure of the irradiation embittlement are shown. As for in-core structures, intergranular stress corrosion cracking is discussed. As for zirconium alloy, the factors affecting the irradiation creep and growth are explained. (K.I.)

  6. 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.

  7. TRIGA Reactor Power Upgrading Analysis

    International Nuclear Information System (INIS)

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

  8. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  9. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

  10. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

  11. 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

  12. 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

  13. TRIGA research reactors with higher power density

    International Nuclear Information System (INIS)

    The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (∼48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D2O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 1014 n/cm2·s and fast fluxes (> 0.1 MeV) of 2 x 1014 n/cm2·s. The core centerline thermal neutron flux in the D2O reflector is about 2 x 1014 n/cm2·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D2O reflector of 3 x 1014 n/cm2·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

  14. 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)

  15. Reactor power for space exploration

    International Nuclear Information System (INIS)

    Potential 21st century missions envisioned by National Aeronautics and Space Administration (NASA) planners encompass ambitious, wide-ranging human and robotic solar system exploration objectives and scenarios. A critical common element in many of these future civil space mission initiatives is the ability to generate, with a very high degree of reliability, the considerable amounts of power needed to realize the mission goals. The extended duration and/or high power level requirements for many missions and, in instances, the lack of adequate solar energy flux for others, render the use of versatile nuclear power sources as either missions-enabling or very advantageous. Further, the use of high-performance reactor systems, when coupled with very high impulse electric propulsion systems, can enable or significantly enhance both human near-planets operations and robotic scientific missions to the very farthest reaches of the solar system. It is important that this nation continue to develop the means of acquiring a space reactor power source to ensure availability at such time that approved missions and possibly political considerations warrant its use

  16. Advancement of light water reactor technology

    International Nuclear Information System (INIS)

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

  17. Fuel assembly identification for nuclear power reactors

    International Nuclear Information System (INIS)

    The standard refers to fuel assemblies of light-water power reactors. It contains stipulations for uniform marking in order that the fuel assemblies may be identified. A figure consisting of 8 alpha-numerical characters is used for marking, the first three of which represent the operator who ordered the fuel assembly, while the four characters to follow symbolize a series number. The last character serves as a test mark to scrutinize reading mistakes. The alpha-numerical characters include the Arabic numerals 0-9 and, following them, the letters A-Y of the German alphabet, leaving out B, F, I, O, Q, Z (30 characters). (orig./HP)

  18. Fuel assembly identification for power reactors

    International Nuclear Information System (INIS)

    The standard refers to fuel assemblies of light-water power reactors. It contains stipulations for uniform marking in order that the fuel assemblies may be identified. A figure consisting of 8 alpha-numerical numbers is used for marking, the first three of which represent the operator who ordered the fuel assembly, while the four numbers to follow symbolize a series number. The last number serves as a test mark to scrutinize reading mistakes. The alpha-numerical numbers include the Arabic numerals 0-9 and, following them, the letters A-Y of the German alphabet, leaving out B, F, I, O, Q, Z (30 characters). (orig./HP)

  19. Prospects for small and medium power reactors

    International Nuclear Information System (INIS)

    A searching examination of the present status of nuclear power technology and economics was made in 64 papers presented to the Conference on Small and Medium Power Reactors held by the IAEA in Vienna during the week 5 - 9 September 1960. The IAEA Conference concentrated on small and medium power reactors because these are the sizes of primary interest to less-developed countries around the world. The Conference brought forward information on a wide range of subjects related to power reactors, including power costs, summaries of national programs, applications in less-developed countries, process heat reactors, reactor safety, results of experience in the actual construction and operation of power reactors and technical appraisals of various reactor types

  20. Feasible reactor power cutback logic development for an integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Han, Soon-Kyoo [KHNP Co., Ltd., Uljin-gun, Gyeong-buk (Korea, Republic of); Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok [Korea Atomic Energy Research Institute (KAERI), Daedeokdaero, Yuseong, Daejeon (Korea, Republic of)

    2013-07-15

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  1. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  2. 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

  3. 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%.

  4. Innovative Pressure Tube Light Water Reactor with Variable Moderator Control

    International Nuclear Information System (INIS)

    The features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide the reactor with considerable flexibility for continuous design improvements and developments. This paper presents the development of innovative pressure tube light water reactor, which has the ability to advance the current pressure tubes reactors. The proposed design is aimed to simplify the pressure tubes reactors by: - replacing heavy water by a light water as a coolant and moderator, - adopting batch refueling instead of on-line refueling. Furthermore, the design is based on proven technologies, existing fuel and structure materials. Therefore, it is reasonable to expect significant capital cost savings, short licensing and introduction period of the proposed concept into the power production grid. The basic novelty of the proposed design is based on an idea of variable moderator content in the core and 'breed and burn' mode of operation. Both concepts were extensively investigated and reported in the past (2) (3) (4). In order to evaluate a practical reactor design build on proven technology, several features of the advanced CANDU reactor (ACR-1000) were adopted. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The proposed design is basically pressure tube light water reactor with variable moderator Control (PTVM LWR). This paper presents a detailed description of the PTVM core design and demonstrates the reactivity control and the 'breed and burn' mode of operation, which are implemented by the variation of the moderator in the core, from a

  5. 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

  6. Light-water reactor accident classification

    Energy Technology Data Exchange (ETDEWEB)

    Washburn, B.W.

    1980-02-01

    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.

  7. Facilitation of decommissioning light water reactors

    International Nuclear Information System (INIS)

    Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented. A wide range of facilitation methods - from improved documentation to special decommissioning tools and techniques - is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given

  8. 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

  9. Light-water-reactor safety research agenda

    International Nuclear Information System (INIS)

    The purpose of this paper is to describe the agenda for light water reactor (LWR) safety research as the author sees it from the perspective of the US Nuclear Regulatory Commission (NRC). This research has a good record of accomplishment in the years since the Three Mile Island Unit 2 (TMI-2) accident in systems, materials, the person-machine interface, and in the knowledge of severe accidents. We are well on the way to sufficiency of knowledge for opeating LWRs, but a few important tasks remain. The advanced LWR concepts arise from a well-established base of design and operating experience, but, in their innovation of concepts and systems, they present some new challenges to designers, regulators, and operators, as the latter approach the time of decision in the newly revised process for review and certification of standard designs. This revised process carries with it a larger commitment in the expected number of plants of the same design over a 15-yr period than was the case when commercial nuclear power began its development in the 1960s

  10. Use of Thorium in Light Water Reactors

    International Nuclear Information System (INIS)

    Thorium-based fuels can be used to reduce concerns related to the proliferation potential and waste disposal of the conventional light water reactor (LWR) uranium fuel cycle. The main sources of proliferation potential and radiotoxicity are the plutonium and higher actinides generated during the burnup of standard LWR fuel. A significant reduction in the quantity and quality of the generated Pu can be achieved by replacing the 238U fertile component of conventional low-enriched uranium fuel by 232Th. Thorium can also be used as a way to manage the growth of plutonium stockpiles by burning plutonium, or achieving a net-zero transuranic production, sustainable recycle scenario. This paper summarizes some of the results of recent studies of the performance of thorium-based fuels.It is concluded that the use of heterogeneous U-Th fuel provides higher neutronic potential than a homogeneous fuel. However, in the former case, the uranium portion of the fuel operates at a higher power density, and care is needed to meet the thermal margins and address the higher-burnup implications. In macroheterogeneous designs, the U-Th fuel can yield reduced spent-fuel volume, toxicity, and decay heat. The main advantage of Pu-Th oxide over mixed oxide is better void reactivity behavior even for undermoderated designs, and increased burnup of Pu

  11. 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

  12. 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

  13. Radioisotope powered light sources

    Energy Technology Data Exchange (ETDEWEB)

    Case, F. N.; Remini, W. C.

    1980-01-01

    Radioisotopes have been used for a number of years to excite phosphors to produce visible light. The advent of the nuclear age, however, made possible the preparation of radionuclides in larger quantities at relatively low prices, and with radiation properties that greatly expanded the potential applications for such lights. Current energy conservation needs and inflation leading to even higher costs for maintenance and capital equipment has provided the incentive for development of illuminators for air field markers using both byproduct krypton-85 and processed tritium. Background and current status of these developments are discussed.

  14. Radioisotope powered light sources

    International Nuclear Information System (INIS)

    Radioisotopes have been used for a number of years to excite phosphors to produce visible light. The advent of the nuclear age, however, made possible the preparation of radionuclides in larger quantities at relatively low prices, and with radiation properties that greatly expanded the potential applications for such lights. Current energy conservation needs and inflation leading to even higher costs for maintenance and capital equipment has provided the incentive for development of illuminators for air field markers using both byproduct krypton-85 and processed tritium. Background and current status of these developments are discussed

  15. Waste disposal from the light water reactor fuel cycle

    International Nuclear Information System (INIS)

    Alternative nuclear fuel cycles for support of light water reactors are described and wastes containing naturally occurring or artificially produced radioactivity reviewed. General principles and objectives in radioactive waste management are outlined, and methods for their practical application to fuel cycle wastes discussed. The paper concentrates upon management of wastes from upgrading processes of uranium hexafluoride manufacture and uranium enrichment, and, to a lesser extent, nuclear power reactor wastes. Some estimates of radiological dose commitments and health effects from nuclear power and fuel cycle wastes have been made for US conditions. These indicate that the major part of the radiological dose arises from uranium mining and milling, operation of nuclear reactors, and spent fuel reprocessing. However, the total dose from the fuel cycle is estimated to be only a small fraction of that from natural background radiation

  16. 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...

  17. 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

  18. Core burnup characteristics of high conversion light water reactor, (1)

    International Nuclear Information System (INIS)

    In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in the reactor (HCLWR-J1) under study. The burnup calculations were carried out under the assumption of three batch and out-in fuel loading from the first cycle to the equilibrium cycle. A detailed evaluation was made for discharge burnup, conversion ratio, power distribution, and reactivity coefficients and so on. (author)

  19. 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

  20. 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.

  1. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    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

  2. 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

  3. 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.

  4. KLT-20 reactor for a floating power unit. Annex VI

    International Nuclear Information System (INIS)

    The KLT-20 reactor installation is being designed by the Experimental Design Bureau of Machine Building (OKBM, Nizhny Novgorod) as a power source for floating nuclear power plants (NPPs). At present, the activities are most advanced for the project of a pilot floating heat and power plant with the KLT-40S reactor installations, advanced analogues of the commercial KLT-40 reactors of the Russian icebreaker fleet. For the KLT-40S, detailed design of the reactor unit and floating power unit has been developed and approved; the Rostechnadzor of Russia license for plant siting and floating power unit construction in Severodvinsk (Russian Federation) has been obtained. The KLT-20, based on a pressurized light-water reactor of 20 MW(e), is a two-loop modification of the KLT-40S reactor with several improvements in the main equipment and a long-refuelling interval, achieved with the enrichment of less than 20%. The reactor design with a long refuelling interval was developed based on the engineering solutions of the pilot KLT-40S reactor installation; different from it, the KLT-20 provides for no on-site refuelling. The refuelling, radioactive waste management and repairs of a floating NPP with the KLT-20 would be performed at special maintenance centres. The infrastructure of nuclear ship maintenance centres in Russia could be used for these purposes

  5. Safety analysis for non-power reactors

    International Nuclear Information System (INIS)

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

  6. 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

  7. Reactor power measuring method and device therefor

    International Nuclear Information System (INIS)

    The present invention concerns measurement of a BWR type reactor power and provides a method of and a device for ensuring accuracy of calibration of sensitivity of neutron detectors and measurement of reactor power even if γ-ray thermometers are failed. Namely, the output signals of the γ-ray thermometers are compared with previously determined judging values to detect failures. The reactor power is measured based on the signals of neutron detectors calibrated by integral thermometers except for neutron detectors calibrated by γ-ray thermometers detected as failed. Calibration for sensitivity of neutron detectors as objects of γ-ray thermometers detected as failed is preferably prohibited. Accuracy of measurement of the reactor power can be ensured by the method described above. If axial power distribution of the reactor core is measured while eliminating the signals of γ-ray thermometers detected as failed, accuracy of the measurement of axial power distribution can be ensured. (N.H.)

  8. Applications of a low power nuclear reactor

    International Nuclear Information System (INIS)

    The Omaha Nebraska Veterans Administration Medical Center (OVAMC) TRIGA reactor is a research reactor designed and fabricated by General Atomic. The reactor first achieved criticality on June 30, 1959. It is a below grade, open-tank-type, ligh water moderated, cooled, and shielded reactor that currently is authorized to operate in the steady-state mode at thermal power levels up to 18KW with an excess reactivity limitation of 0.79% Delta K/K. (author)

  9. 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

  10. 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

  11. The Consortium for Advanced Simulation of Light Water Reactors

    International Nuclear Information System (INIS)

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  12. The Consortium for Advanced Simulation of Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  13. 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.

  14. Automatic power control system for 235 MWe atomic power reactor

    International Nuclear Information System (INIS)

    The paper highlights the essential features of the design, fabrication and testing of microprocessor based reactor power regulating system of Narora Atomic Power Plant (NAPP) and Kakrapar Atomic Power Plant (KAPP). The improved system design at KAPP employs the reactor power control based on neutron flux signal after correction. The control system responses have been presented and compared with the responses using a reactor functional simulator. A new fault tolerant reactor regulating system has been designed using a dual active and hot stand-by microprocessor system to improve operational reliability. (author). 1 ref., 8 figs

  15. 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.

  16. 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)

  17. 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.)

  18. 137Cs dynamics within a reactor effluent stream in South Carolina

    International Nuclear Information System (INIS)

    Radiocesium dynamics are being studied in a blackwater creek which had received production reactor releases from the Savannah River Plant in South Carolina. Most 137Cs in the water column is dissolved or in colloidal form and is believed to originate primarily through outflow from an upstream ''contaminated'' reservoir. All ecosystem components in the stream have high 137Cs concentration factors. Radiocesium concentrations are highest in filamentous algae (332 pCi/g-dry) and suspended particulate matter (100 to 200 pCi/g). Other food chain bases had much lower 137Cs levels. Most consumer populations averaged 10 to 50 pCi/g. Radiocesium concentrations decreased in transfers between food chain bases and primary consumers or filter feeders. Omnivores and small predators have similar 137Cs concentrations with bioaccumulation occurring by top-carnivores. Radiocesium levels are around 100 pCi/g in largemouth bass and water snakes. Foodweb components in the stream have reached a dynamic equilibrium in 137Cs concentrations despite a 10 yr absence of reactor operations. Radiocesium levels are apparently being maintained through long-term 137Cs cycling in the upstream reservoir and surrounding flood plain forest systems. Rainfall and other physical processes influence the seasonal 137Cs fluctuations in stream components

  19. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  20. Computer code for thermal hydraulic analysis of light water reactors

    International Nuclear Information System (INIS)

    A computer programme (THAL) has been developed to perform thermal hydraulic analysis of a single channel in a light water moderated core. In this code the hydrodynamic and thermodynamic equations describing one-dimensional axial flow have been discretized and solved explicitly stepwise up the coolant channel for an arbitrary power profile. THAL has been developed for use on small computers and it is capable of predicting the coolant, clad and fuel temperature profiles, steam quality, void fraction, pressure drop, critical heat flux and DNB ratio throughout the core. A boiling water reactor and a pressurized water reactor have been analyzed as test cases. The results obtained through the use of THAL compare favourably with those given in the design reports of these reactor systems. (author)

  1. Incorporating outage management principles into the advanced light water reactor

    International Nuclear Information System (INIS)

    In the United States there are 110 light water reactor (LWR) plants currently in operation, with a total generating capacity of 102 580 MW(electric). These plants include 37 boiling water reactor (BWR) and 73 pressurized water reactor (PWR) units. Since 1980, more than 40 nuclear power plants have entered service in the United States. However, no new plants have been ordered by utilities and owners groups since 1978. There will come a time in the not-too-distant future that new, large electricity generating units will be needed to supply expected increases in base-load capacity. Will the new advanced LWR (ALWR) designs be able to pass muster and be chosen to help meet that need? With outage management at operating plants improving every year, what can the ALWR designs offer that has not already been incorporated?

  2. 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

  3. 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.)

  4. 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.)

  5. 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

  6. 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

  7. Thyristor Controlled Reactor for Power Factor Improvement

    Directory of Open Access Journals (Sweden)

    Sheila Mahapatra

    2014-04-01

    Full Text Available Power factor improvement is the essence of any power sector for reliable operation. This paper provides Thyristor Controlled Reactor regulated by programmed microcontroller which aids in improving power factor and retaining it close to unity under various loading conditions. The implementation is done on 8051 microcontrollerwhich isprogrammed using Keil software. To determine time lag between current and voltage PSpice softwareis used and to display power factor according tothe variation in loadProteus software is used. Whenever a capacitive load is connected to the transmission linea shunt reactor is connected which injects lagging reactive VARs to the power system. As a result the power factor is improved. The results given in this paper provides suitable microcontroller based reactive power compensation and power factor improvement technique using a Thyristor Controlled Reactor module.

  8. 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

  9. 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

  10. 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

  11. 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)

  12. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

  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. Qualification issues for advanced light-water reactor protection systems

    International Nuclear Information System (INIS)

    The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying the I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light-water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light-water reactor. The template was then used to suggest a methodology for the qualification of microprocessor-based protection systems. The methodology identifies standards/regulatory guides (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems. This approach addresses in part issues raised in NRC policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards. as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

  15. Development of light water reactors and cooling systems

    International Nuclear Information System (INIS)

    Development of Light Water Reactors (LWR) started in US. Japan imported this technology from US, and its construction and improvement had been done by adding Japanese original technologies. This article outlined development history of LWR, its plant system and main components, ECCS and accident management. Most operating LWR were those rapidly developed from 1960 to 1970 and associated accident response was so designed for all assumed conditions at that time. At the Fukushima Daiichi nuclear power station, needed works could not be down well due to all losses of AC and DC power, inability to assure plant state at control room under power stoppage, work interruption caused by aftershock and road blockage by Tsunami drift, and accident management was not effective and accident was enlarged. After the Fukushima nuclear accident, enhanced measures against more stringent conditions such as external events should be prepared to assure safety of nuclear power station using latest knowledge. (T. Tanaka)

  16. 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

  17. 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

  18. 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

  19. 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)

  20. reactor power control using fuzzy logic

    International Nuclear Information System (INIS)

    power stabilization is a critical issue in nuclear reactors. convention pd- controller is currently used in egypt second testing research reactor (ETRR-2). two fuzzy controllers are proposed to control the reactor power of ETRR-2 reactor. the design of the first one is based on a set of linguistic rules that were adopted from the human operators experience. after off-line fuzzy computations, the controller is a lookup table, and thus, real time controller is achieved. comparing this f lc response with the pd-controller response, which already exists in the system, through studying the expected transients during the normal operation of ETRR-2 reactor, the simulation results show that, fl s has the better response, the second controller is adaptive fuzzy controller, which is proposed to deal with system non-linearity . The simulation results show that the proposed adaptive fuzzy controller gives a better integral square error (i se) index than the existing conventional od controller

  1. 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

  2. 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

  3. BN-1200 reactor power unit design development

    International Nuclear Information System (INIS)

    In February 2010, the RF Government has approved the Federal Target Program “New Generation Nuclear Power Technologies for the Period of 2010-2015 and for the long-term up to 2020”. Within this Program, the R&D work for new generation 1200MWe sodium fast reactor is provided. The BN-1200 design is based on the combination of approved and innovative technical decisions, which allow: – reliable power unit with large BN reactor to be developed in a short period of time for commercial construction as a part of closed nuclear fuel cycle; – qualitatively new technical level of power unit to be provided according to generation 4 NPP requirements. The paper characterizes the activities performed now for the power unit design in various areas: – power unit design; – reactor plant (RP) detailed design development; – R&D work to validate the RP system and equipment; – code upgrading and verification; – safety validation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-04

    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.

  5. 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

  6. Preliminary concepts: safeguards for spent light-water reactor fuels

    International Nuclear Information System (INIS)

    The technology available for safeguarding spent nuclear fuels from light-water power reactors is reviewed, and preliminary concepts for a spent-fuel safeguards system are presented. Essential elements of a spent-fuel safeguards system are infrequent on-site inspections, containment and surveillance systems to assure the integrity of stored fuel between inspections, and nondestructive measurements of the fuel assemblies. Key safeguards research and development activities necessary to implement such a system are identified. These activities include the development of tamper-indicating fuel-assembly identification systems and the design and development of nondestructive spent-fuel measurement systems

  7. ALWR - VTT's technology programme on advanced light water reactors

    International Nuclear Information System (INIS)

    In January 1998, VTT Energy launched a four year research programme 'Advanced Light Water Reactors' (ALWR). The research programme strives to intensify co operation between different Finnish organisations as well as to make better use of international R and D efforts on ALWR concepts. Some projects of the programme focus on use of new computational tools for the design and safety analysis of ALWRs. Improved computer simulation methods and models are necessary for the analysis of passive safety systems. Some projects address development and assessment of new technical solutions in ALWR. The projects are carried out in close co operation with the Finnish power companies and are often part of reactor vendors international development programmes. Another key element of the programme is training of new nuclear experts and providing possibilities for continuing education of the current staff. (author)

  8. 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

  9. Power calibration study at the Musashi reactor

    International Nuclear Information System (INIS)

    The Musashi reactor (TRIGA-II,100 kW) initially went critical in January of 1963. The reactor had been used for training, isotope production and medical irradiation for boron neutron capture therapy (1). The initial power calibration was based on the use of a calibrated electrical heater in a calorimetric procedure where the rate of rise of the bulk pit water temperature was measured using 2 kW heaters x 6 pieces. The rate of rise of water temperature was determined to be 0.0474 C/kWh. The reactor was then operated to give the same rate of rise of water temperature. Thus the reactor power was established at the value produced by the electrical heaters. A stirrer for tank water mixing was not used. Recent communications (2)(3) indicated that power calibrations using a stirrer provided a much more uniform mixing, and heating in the reactor tank water which was essential for an accurate calibration. In this paper, the effect of mixing using a stirrer was investigated considering the physical factors such as room temperature, humidity, tank water temperature and it's distributions. The room temperature and humidity around the reactor varies 6-30 and 30-80 %, respectively, depending on four seasons. The heat flow through the surface of the pool was also evaluated because the reactor usually operates without cover on the surface of the pool. (orig.)

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. The advanced light water reactor program approach for gaining acceptance

    International Nuclear Information System (INIS)

    Electric utilities in the US face several obstacles and disincentives to building new nuclear power plants. A reformed licensing process that permits true one-step licensing is a prerequisite, as is timely implementation of the technical and political solutions of the high-level waste issue by the federal government and low-level waste storage by state governments. In addition, a more tangible acceptance of the need, benefits, and residual risk of nuclear power by all concerned institutions is an essential impetus for the return of the nuclear option. An improved version of the light water reactor (LWR) is expected to be the preferred choice for the next increment of nuclear capacity ordered in the US. The paper discusses the need for research and advanced LWR development. The advanced LWR program, sponsored jointly by the utility industry and the US Department of Energy (DOE), offers the most promising approach to developing the next generation of nuclear power plants

  16. 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)

  17. 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) 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

  18. 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

  19. Improvement of the axial power distribution control capabilities in WWER-1000 reactors

    International Nuclear Information System (INIS)

    This article describes an automatic reactor power control system for WWER-1000 reactor and reports simulation analysis results for a typical daily load follow operation. The associated reactor control algorithm is called 'mode G' that uses a heavy-worth bank (H-bank) dedicated to axial power shape control and the light-gray banks (G-banks) for reactor power change and reactivity compensation. The simulation results for daily load follow operation in three burnup states of first cycle illustrate that the load follow capability of WWER-1000 reactors using this algorithm will be improved

  20. NEPTUNE: a modular system for light-water reactor calculation

    International Nuclear Information System (INIS)

    A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE

  1. 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.

  2. 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.)

  3. 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

  4. 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.

  5. 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.

  6. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [INL; Busby, Jeremy [ORNL; Hallbert, Bruce [INL; Bragg-Sitton, Shannon [INL; Smith, Curtis [INL; Barnard, Cathy [INL

    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.

  7. The Swedish Zero Power Reactor R0

    International Nuclear Information System (INIS)

    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

  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. 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.

  10. 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

    International Nuclear Information System (INIS)

    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 Plantsclose 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 ampersand 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 ampersand 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

  11. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  12. 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

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

    International Nuclear Information System (INIS)

    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

  14. 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.

  15. 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.

  16. CNSC power reactor operating licence reform

    International Nuclear Information System (INIS)

    CNSC staff introduced a new Power Reactor Operating Licence (PROL) in order to strengthen the regulatory oversight of power reactor operation, while increasing regulatory effectiveness and efficiency by focusing on risk-significant issues and reducing purely administrative efforts. The PROLs have been simplified by incorporating a more risk-informed approach and by eliminating cascading references to working level licensee documentation and regulatory expectations. To ensure that there is a common understanding for each requirement specified in the PROL, CNSC staff prepared a Licence Conditions Handbook (LCH), which provides technical details and compliance verification criteria on how licence conditions are to be met. (author)

  17. Reactor power system deployment and startup

    Science.gov (United States)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. Current status of light water reactor and Hitachi's technical improvements for BWR

    International Nuclear Information System (INIS)

    Gradual technical improvements in Japan over the years has improved the reliability of light water reactors, and has achieved the highest capacity factor level in the world. Commercial operation of Fukushima 2-2 (1,100 MW) of the Tokyo Electric Power Co. was started in February, 1984, as the first standardized BWR base plant, ushering in a new age of domestic light water reactor technology. The ABWR (1,300 MW class) has been developed as Japan's next generation light water reactor, with construction aimed at the latter half of the 1980's. Hitachi's extensive efforts range from key nuclear equipment to various related robots, directed at improving safety, reliability, and the capacity factor, while reducing radiation exposure. This paper presents an outline of Hitachi's participation in the light water reactor's improvement and standardization, and the current status of our role in the international cooperation plan for the ABWR. (author)

  4. 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

  5. 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)

  6. 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

  7. Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

    International Nuclear Information System (INIS)

    Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 233U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

  8. 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

  9. 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

  10. Solid state track recorder pressure vessel surveillance neutron dosimetry at commercial nuclear power reactors

    International Nuclear Information System (INIS)

    Solid State Track Recorder neutron dosimetry methods developed under the U.S. Nuclear Regulatory Commission Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program have been applied for pressure vessel surveillance dosimetry at commercial nuclear power reactors. More than 800 SSTR neutron dosimeters have been deployed at twelve different power reactors during twenty-two cycles of operation. More than 300 SSTR have been analyzed, and results with uncertainties in the 2-5% range have been generally obtained. Several new areas of application of SSTRs for radiation damage assessment for safe routine operation or extended life operation of power reactors are planned and these applications are discussed. (author). 14 refs

  11. Status of fusion power reactor design activity in JAERI

    International Nuclear Information System (INIS)

    A number of fusion power reactor design studies conducted in JAERI over the past 22 years are reviewed and the present status of reactor studies is introduced. A helium gas cooled power reactor which has a blanket with solid lithium ceramics such as Li2O for the breeding material and incolloy-800 for the structural material was proposed in 1973. This is one of the first reactor design using a lithium ceramics blanket concept. Another power reactor design called the Swimming Pool-type Tokamak Reactor (SPTR-P) was completed in 1983. In SPTR-P, the tokamak reactor is submerged in a water pool to utilize water as the shield to reduce long term radioactive waste and to achieve easy repair and maintenance. In 1990, the Steady State Tokamak Reactor (SSTR) concept was proposed as a realistic fusion reactor to be built in the near future. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required to maintain steady state operation. It is cooled by pressurized water and uses low activation ferritic steel as the structural material. A study of SSTR-2 to improve the safety and economic aspects of SSTR by replacing the water coolant by a mixture of helium gas and fine solid particles was carried out in 1992. Light yet highly heat resistant material, titanium aluminide intermetallic compound (TiAl), is used as the structural material. The net thermal efficiency larger than 40 % can be achieved with the gas-particulate mixture at 5 MPa and the outlet coolant temperature of 700degC. A concept of a Drastically Easy Maintenance (DREAM) tokamak reactor has recently been proposed. For easy replacement of blanket and divertor, a plasma configuration with high aspect ratio around 6 and a small number of torus sectoring of 12 are selected. A 1/12 torus sector is horizontally pulled out between the TF coils with a straight radial motion. It is estimated that the availability of 85 % can be achieved. (author)

  12. Sustainability of light water reactor fuel cycles

    International Nuclear Information System (INIS)

    This paper compares the sustainability of two light water reactor, LWR, fuel cycles: the once-through UOX (low-enriched uranium oxide) cycle and the twice-through MOX (Mixed Uranium-Plutonium Oxide) cycle (increasing the input efficiency of available uranium) by assessing their probable long-term competitiveness. With the retirement of diffusion enrichment facilities, enrichment prices have declined by one-third since 2009 and are likely to remain below $100-kgSWU for the foreseeable future. Here, initial uranium prices are set at $90/kgU and reprocessing costs at $2500 per kilogram of heavy-metal throughput, representative of “new-build” costs for reprocessing facilities. Substantial reprocessing cost reductions must be achieved if MOX is to be competitive, i.e., if it is to improve the sustainability of the LWR. However, results indicate that preserving the MOX alternative for spent fuel management later in this century has a large present value under several sets of assumptions regarding uranium price increases and reprocessing cost decreases. - Highlights: • We compare two nuclear fuel cycles: uranium versus reprocessed plutonium (mixed oxide, MOX). • We modify assumptions in MIT, The Future of the Nuclear Fuel Cycle (2011). • New reprocessing facilities are more expensive and uranium enrichment prices are lower than previously assumed, decreasing MOX competitiveness. • The MOX cycle option value could be large, but depends on uranium prices and reprocessing costs. • R and D should focus on reducing reprocessing facility costs before implementing the MOX fuel cycle

  13. Reactor power distribution pattern judging device

    International Nuclear Information System (INIS)

    The judging device of the present invention comprises a power distribution readout system for intaking a power value from a fuel segment, a neural network having an experience learning function for receiving a power distribution value as an input variant, mapping it into a desirable property and self-organizing the map, and a learning date base storing a plurality of learnt samples. The read power distribution is classified depending on the similarity thereof with any one of representative learnt power distribution, and the corresponding state of the reactor core is outputted as a result of the judgement. When an error is found in the classified judging operation, erroneous cases are additionally learnt by using the experience and learning function, thereby improving the accuracy of the reactor core characteristic estimation operation. Since the device is mainly based on the neural network having a self-learning function and a pattern classification and judging function, a judging device having a human's intuitive pattern recognition performance and a pattern experience and learning performance is obtainable, thereby enabling to judge the state of the reactor core accurately. (N.H.)

  14. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  15. 78 FR 8190 - Commercial Leasing for Wind Power on the Outer Continental Shelf (OCS) Offshore North Carolina...

    Science.gov (United States)

    2013-02-05

    ... 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 in... Bureau of Ocean Energy Management Commercial Leasing for Wind Power on the Outer Continental Shelf...

  16. Utilization of stable isotopes in power reactor

    International Nuclear Information System (INIS)

    The stable isotopes, besides uranium, used in EDF power nuclear reactors are mainly the boron 10 and the lithium 7. Boron is used in reactors as a neutrophagous agent for core reactivity control, and lithium, and more especially lithium 7, is extensively used as a solution in PWR moderators for primary fluid pH control. Boron and lithium ore reserves and producers are presented; industrial isotopic separation techniques are described: for the boron 10, they include dissociative distillation (Sulzer process) and separation on anionic resins, and for lithium 7, ion exchange columns (Cogema). 1 tab

  17. Small power sodium cooled fast nuclear reactors

    International Nuclear Information System (INIS)

    1.5 MW(e), 12 MW(e) and 170 MW(e) small power sodium cooled fast reactors have been developed. The reactor plants were developed as universal power units for economically effective energy and industrial steam generation and heat supply. The main features increasing the power unit economic efficiency are: serial fabrication of standard RPs at the factory and delivery of reactor vessels in ready made form; realization of self-protection principles and use of passive systems in RP; use of standard machine room equipment, fabricated in accordance with the rules of conventional heat power engineering; use of turbine plant with thermodynamic coefficient, exceeding the corresponding value for the plants of PWR type. For MBRU-1.5 and MBRU-12 RPs it is proposed to use a core without FA replacement during the whole service life (30 years) and for BMN-170 RP it is proposed to use a core with a 4 year operating period and 1 year between the refueling shutdowns. During the whole service life a minimal number of operating personnel will be needed for the plant servicing. The personnel functions will be periodically to observe the parameters of technological process. Passive principles are used in the main RP safety systems: a passive type system of emergency residual heat removal system provides heat removal directly through the reactor vessel forced air cooling due to the natural air chimney effect; an emergency reactor shut-down system is provided by emergency protection rods with active-passive action. (author)

  18. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  19. 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

  20. 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.

  1. 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)

  2. History and status of reloading techniques for light water reactors

    International Nuclear Information System (INIS)

    The history and status of reloading techniques for light water reactors is presented by covering both the techniques used to reload the cores and the nucleonic codes used to calculate the core performance characteristics. The nucleonic codes involve a cross section generating code and a core analysis code. The evolution of these codes is described beginning with the early LEOPARD code and then going to present day WIMS, CASMO, CPM2, etc. for the cross section generating codes. The modern accurate nodal codes with pin power reconstruction is presented as the most advanced method for analyzing reloads. The basic equations employed in modern cross section generating codes and modern nodal codes are developed. Optimization techniques used to reload nuclear reactors are briefly reviewed, i.e., giving a short history on this subject and ending with modern optimization techniques. Special emphasis is given to the author's method of using one code system to perform automatically all of the tasks needed to reload the reactor. Two separate codes are described, one involving a combination of C-language and FORTRAN, and the other completely written in FORTRAN. The former is designed for the Westinghouse Beaver Valley PWR's and the other for the TIM-1 PWR. Each code contains the optimization calculations needed to produce an optimal reload design with all of the numerous constraints needed for a practical core. Both codes, however, employ the same basic approach. They begin with a priority loading plan (its source described in the paper) which initially loads the core in an optimum manner with the available fuel. The core pattern is then automatically modified to meet all constraints using the Haling power distribution. Then the burnable poisons are optimally placed in the core. The final design is automatically depleted in the normal manner. These codes can load a core equal to or better than any experienced fuel manager. (author). 43 refs, 32 figs, 2 tabs

  3. 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

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

    International Nuclear Information System (INIS)

    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. 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...

  6. 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...

  7. 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

  8. 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

  9. Light Water Reactor Generic Safety Issues Database (LWRGSIDB). User's manual

    International Nuclear Information System (INIS)

    resolved in other plants and which can be used in reassessing the safety of individual operating plants. The IAEA-TECDOC-1044, Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for Their Resolution (September 1998), is a compilation of such safety issues which is based on broad international experience. This compilation is one element within the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It is a compilation not only of the generic safety issues identified in nuclear power plants but also, in almost all cases, the measures taken to resolve these issues. The safety issues, which are generic in nature with regard to light water reactors (LWRs), and the measures taken for their resolution, are intended for use as a reference in the reassessment of the safety of operating plants.The information contained in the main body of the TECDOC has been used to establish a database. This database has search, query and report functions. This information is thus available in an electronic form which can be selectively queried and with which reports can be produced according to the requirements of the user. The database also enables the IAEA to update the data periodically on the basis of information made available by Member States

  10. 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)

  11. 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.

  12. 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

  13. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author)

  14. 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

  15. Enhancement of the quality of the reactor pressure vessel used in light water power plants by advanced material, fabrication and testing technologies

    International Nuclear Information System (INIS)

    Fracture safe assessment of nuclear reactor pressure vessels (RPV) is based upon an adequate stress analysis, reliable material characteristics, and acceptable defect sizes. Problems may arise concerning inhomogeneties, low toughness and crack phenomena as observed in the base material and heat affected zone (HAZ). Therefore, efforts have been made to develop a steel which would be both non-susceptible to embrittlement and/or cracking in the HAZ, and have a higher upper-shelf toughness of base and HAZ material. Tests have been made on inhomogeneties and defects and also on improvement of chemical composition, the steel-making process, welding procedures and the optimum temperature cycle and level for stress-relief heat treatment. To solve these problems, common testing methods were supplemented by tangential-cut techniques, small HAZ-tensile test procedures and HAZ-simulation techniques. Results indicate that 50 per cent of 100 investigated component-strength welds are affected by micro stress-relief cracking (SRC) on a micro-and millimetre scale. The 22 NiMoCr 37 steel with optimised chemical composition, and the 20 MnMoNi 55 steel are both resistant to stress-relief embrittlement and SRC. Specific welding techniques are found to limit SRC and proposals for optimum stress-relief temperatures are given. For the generation of new components, the fracture-safe analysis can now be based completely upon homogeneous and high upper-shelf base materials including the HAZ. (author)

  16. Radioactive wastes from possible future UK Light Water Reactors

    International Nuclear Information System (INIS)

    The predictive radioactive wastes from the operation and decommissioning of various designs of Light Water Reactor (LWR) are reviewed and compared to the Sizewell 'B' Pressurised Water Reactor. The designs considered are the N4, Konvoi, System 80+, Advanced PWR, AP600, Advanced Boiling Water Reactor, the European Pressurised Water Reactor (EPR) and the Sizewell 'XL' design. The sources of activity, active arisings, waste treatment plants, discharges to the environment, the optimisation of waste treatment, lifetime arisings of solid waste and the characteristics of spent fuel are addressed. Conclusions are drawn on the implications for UK waste management policy. (author)

  17. 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).

  18. 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

  19. POWERED LED LIGHTING SUPPLIED FROM PV CELLS

    OpenAIRE

    Tirshu M.; Uzun M; Speian A.; Spivac V.; Bogdan A.

    2011-01-01

    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 me...

  20. High performance light water reactor core design studies

    International Nuclear Information System (INIS)

    The High Performance Light Water Reactor (HPLWR) is the European version of the various supercritical water cooled reactor proposals. The paper presents the activity of KFKI-AEKI in the field of neutronic core design within the framework of the European funded project: 'HPLWR Phase 2'. As the coolant density along the axial direction shows remarkable change, coupled neutronic-thermohydraulic calculations are essential which take into account the heating of moderator in the special water rods of the assemblies. A parametrized diffusion cross section library was prepared for the HPLWR assembly with the MULTICELL neutronic transport code. The parametrized cross sections are used by the KARATE program system, which was verified by comparative Monte Carlo calculations. Methodical core calculations have been carried out. Without thermal insulation in the second superheater of the three pass core, the temperature of moderator water exceeds the pseudocritical temperature resulting in poor moderation. Applying heat insulators helps to avoid the extreme heat up of moderator, so it provides better moderation for neutrons. The increase of the local multiplication in the superheaters favorably affects the power distribution and keff of the core. Preliminary loadings of the HPLWR core were assessed, which contain insulated assemblies with Gd burnable absorbers. The initial reactivity excess was compensated by control rods. The reactivity coefficients of HPLWR with respect to temperatures show substantial negative feedback at normal operating conditions. (author)

  1. 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)

  2. Power maximization of a spheric reflected reactor with optimized fuel distribution

    International Nuclear Information System (INIS)

    The maximum power of a spheric reflected reactor was determined using the theory of optimal control. The control variable employed was the fuel distribution, in accordance to constraints on the power density and on the concentration fuel. It was considered a thermal reactor with a fixed radius. The reactor was fuelled with U-235 and moderated with light water. The nuclear reactor was described by a diffusion theory model. The analytical solution was obtained for both two and four groups of energy and a FORTRAN program was developed to obtain the numerical results. (author)

  3. NUSCALE (NuScale Power, Inc., USA) [Passive Safety Systems in Advanced Small Modular Reactors

    International Nuclear Information System (INIS)

    A NuScale plant consists of 1 to 12 independent modules, each capable of producing a net electric power of 45 MW(e). Each module includes a Pressurized Light Water Reactor operated under natural circulation primary flow conditions. Each reactor is housed within its own high pressure containment vessel which is submerged underwater in a stainless steel lined concrete pool. The cross-sectional view of NuScale reactor building is shown

  4. 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)

  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. 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

  7. 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

  8. Operating U.S. power reactors

    International Nuclear Information System (INIS)

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the United States which have been issued operating licenses. Table I shows the number of such reactors and their net capacities as of June 30, 1992, the end of the three-month period covered in this report Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three-months covered in each report and the cumulative values of these parameters at the end of the covered quarter since the beginning of commercial operation. The information for this table was obtained from the Nuclear Regulatory Commission (NRC) Office of Information Resources Management. The Maximum Dependable Capacity (MDC) Unit Capacity (in percent) is defined as follows: (Net electrical energy generated during the reporting period x 100) divided by the product of the number of hours in the reporting period and the MDC of the reactor in question. The forced outage rate (in percent) is defined as: (The total number of hours in the reporting period during which the unit was inoperable as the result of a forced outage x 100) divided by the sum (forced outage hours + operating hours)

  9. Operating U.S. power reactors

    International Nuclear Information System (INIS)

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the United States which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of Sept. 30, 1990, the end of the three-month period covered in this report. Table 2 used to list the unit capacity and forced outage rate for each licensed reactor for each of the three months covered in each report and the cummulative values of these parameters since the beginning of commercial operation. Since, however, Nuclear Regulatory Commission (NRC) publication NUREG-0200, known as 'The Gray Book,' which has been the source of this information, has been discontinued, in this issue, and henceforth until other data again become available from the NRC, this table will contain maximum dependable capacity (MDC) factors for each month and the year-to-date MDC capacity as of the end of the quarter covered, in this issue the end of September 1990. The information for this table is derived from World Nuclear Performance, published by McGraw-Hill Nuclear Publications and used with their permission. The MDC Unit Capacity (in percent) is defined as follows: (Net electrical energy generated during the reporting period x 100) divided by the product of the number of hours in the reporting period and the MDC of the reactor in question

  10. Fuel cycle options for light water reactors in Germany

    International Nuclear Information System (INIS)

    In Germany 19 nuclear power plants with an electrical output of 22 GWe are in operation. Annually about 450 t of spent fuel are unloaded from the reactors. Currently most of the spent fuel elements are shipped to France and the United Kingdom for reprocessing according to contracts which have been signed since the late 70es. By the amendment of the Atomic Energy Act in 1994 the previous priority for reprocessing of spent nuclear fuel was substituted by a legal equivalency of the reprocessing and direct disposal option. As a consequence some utilities take into consideration the direct disposal of their spent fuel for economical reasons. The separated plutonium will be recycled as MOX fuel in light water reactors. About 30 tons of fissile plutonium will be available to German utilities for recycling by the year 2000. Twelve German reactors are already licensed for the use of MOX fuel, five others have applied for MOX use. Eight reactors are currently using MOX fuel or used it in the past. The spent fuel elements which shall be disposed of without reprocessing will be stored in two interim dry storage facilities at Gorleben and Ahaus. The storage capacities are 3800 and 4200 tHM, respectively. The Gorleben salt dome is currently investigated to prove its suitability as a repository for high level radioactive waste, either in a vitrified form or as conditioned spent fuel. The future development of the nuclear fuel cycle and radioactive waste management depends on the future role of nuclear energy in Germany. According to estimations of the German utilities no additional nuclear power plants are needed in the near future. Around the middle of the next decade it will have to be decided whether existing plants should be substituted by new ones. For the foreseeable time German utilities are interested in a highly flexible approach to the nuclear fuel cycle and waste management keeping open both spent fuel management options: the closed fuel cycle and direct disposal of

  11. 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.

  12. Nuclear power in Kazakhstan and current status of the BN-350 fast reactor

    International Nuclear Information System (INIS)

    Atomic scientific-industrial complex of the Republic of Kazakhstan consists of: Uranium mining, production and power industry which includes enterprises of uranium ores geological searching and a number of natural mines (using the mining and underground leaching techniques); two plants of U3O8 production at the towns Aktau and Stepnogorsk; metallurgical plant producing uranium fuel pellets for fuel assemblies of RBMK and WWER reactors types; energy plant at Aktau (MAEK) is used for production of heat, electricity and desalination of water and based on three energy blocks using natural gas and one nuclear unit with fast breeder reactor BN-350. The fast breeder reactor BN-350 at Aktau was commissioned in November 1972 and finally stopped in April 1999. Three different types of the research reactors on the territory of the former Semipalatinsk Nuclear Test Site and one research reactor and sub critical assembly nearly Almaty are exploited for the investigation in field of reactors nuclear safety and other type of investigations. These are: WWR-K - light water reactor, power - 10 MW; EWG-1M thermal light water heterogeneous vessel reactor with light water moderator and coolant, beryllium reflector, maximum thermal power - 35 MW; RA - thermal neutron high temperature gas heterogeneous reactor with air coolant, zirconium hydride moderator, and beryllium reflector. A brief description of the project on BN-350 spent fuel storage is included with the calculations on safety validity during packaging and storage of the spent fuel elements

  13. 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).

  14. 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.

  15. 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.

  16. Study of power reactor dynamics by stochastic reactor oscillator method

    International Nuclear Information System (INIS)

    Stochastic reactor oscillator and cross correlation method were used for determining reactor dynamics characteristics. Experimental equipment, fast reactor oscillator (BOR-1) was activated by random pulses from the GBS-16 generator. Tape recorder AMPEX-SF-300 and data acquisition tool registered reactor response to perturbations having different frequencies. Reactor response and activation signals were cross correlated by digital computer for different positions of stochastic oscillator and ionization chamber

  17. Utilization of the low power Musashi reactor

    International Nuclear Information System (INIS)

    Although the Musashi reactor is a low-power reactor of 100 kW, multi-purpose beam-experiments have been performed for the last ten years. Medical irradiation for boron neutron capture therapy (BNCT) is the most unique utilization of the reactor. Eighty-two patients had been treated in the reactor up to the end of August 1987. One of the horizontal beam ports has been used for a time-of-flight experiment by using a slow-chopper since 1977. The authors measured the total neutron cross sections of Mg, Al, Si, Zr, Nb and Mo in the energy range from 0.001 to 0.3 eV. A neutron radiography facility was designed and installed at another beam port in 1984. A real-time neutron TV system has been also installed for investigation of moving objects and for a neutron computed tomography study. A third beam port has been used for a filtered beam experiment and a capture γ-ray measurement. An Fe-filter for 24 keV neutrons and Si-filter for 54 and 144 keV neutrons are available for generating monochromatic neutrons. These beams have been used for the precise measurement of total neutron cross sections. The capture γ-ray measurements have been applied for the measurement of boron concentration in tissue in connection with the BNCT. The reactor has a Joint Use Program for university researchers in Japan under a grant-in-aid by the Ministry of Education, Science and Culture. (author)

  18. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  19. Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 2

    International Nuclear Information System (INIS)

    This report, Supplement No. 2 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 No. 1

  20. Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 4

    International Nuclear Information System (INIS)

    This report, Supplement No. 4 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, 2, and 3

  1. 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

  2. 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

  3. 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

  4. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, 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. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

  5. Review of the use and state of development of the various reactor types

    International Nuclear Information System (INIS)

    The report gives a review of the reactor types being of importance from today's point of view for use as stationary power reactors. These are heavy water reactors, light water reactors (pressurized water reactor, Soviet pressurized water reactor, Soviet light-water-graphite reactors, boiling water reactors), gas-cooled reactors (gas-graphite reactors, high temperature reactors), and fast breeder reactors. (HJ)

  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 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

  7. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10-8 cm2/s for boric acid waste form and 9 x 10-9 cm2/s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10-11 cm2/s and 9 x 10-11 cm2/s respectively. (Author)

  8. 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...

  9. Prospects of small light-water reactor utilization

    International Nuclear Information System (INIS)

    OKB Mechanical Engineering (OKBM) has accumulated a many years of experience of designing world famous low power PWRs, including the KLT-40 type propulsion reactors and has gained from the positive experience of their operation and technical perfection. With reliance on this experience, a whole range of prospective small power reactor plants with KLT-40S-type reactors has been developed, including ABV, ATEC co-generation power plants, SAKHA etc., which may be recommended for the immediate future as small power nuclear sources of energy for regions with a decentralized power supply, where the utilization of traditional Diesel generators in entails, in the majority of cases, considerable difficulties associated with the need for stable provision of large quantities of liquid fuel and lubricants. The report contains information on the peculiarities of small power RPs utilization. We also consider the major characteristics of nuclear power plants based on the advanced reactors of the above mentioned types, which meet all modern safety requirements. We demonstrate the possibility of the use of small power RPs in complexes for electric power, heat and fresh water co-generation. (author)

  10. Thermal core design of the high performance light water reactor

    International Nuclear Information System (INIS)

    The High Performance Light Water Reactor (HPLWR) is a SCWR concept, operated at an inlet pressure of 25 MPa with a core outlet temperature of 500 deg. C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 deg. C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Each fuel assembly is built from 40 fuel pins with 8 mm diameter and a pitch of 9.4mm, housed in a thermally insulated assembly box. Additional moderator water is foreseen in water rods inside each assembly and in gaps between the assembly boxes. With a thermal power of 2300 MW, a net electric power of 1000 MW shall be achieved, resulting in a net efficiency of 43.5%. This concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility, which will be summarized in this paper. Coupled neutronic / thermal-hydraulic analyses by Maraczy et al. with the 2 group diffusion code KARATE and the one-dimensional code SPROD are defining an initial distribution of fuel enrichment, the positioning of the control rods, and the use of the burnable Gd absorbers to reach the envisaged power distribution. An equilibrium cycle analysis is showing radial form factors and the discharge burn-up. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. Sub-channel analyses by Himmel et al. performed for different radial power gradients demonstrate the excellent coolant mixing inside assemblies thanks to the wire wrap spacers used in this design. Coolant mixing above and underneath the core has been studied by Wank et al

  11. Nuclear Power Reactors in the World. 2010 Edition

    International Nuclear Information System (INIS)

    This is the thirtieth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: - General and technical information as of the end of 2009 on power reactors operating or under construction, and shut down; - Performance data on reactors operating in the Agency's Member States, as reported to the IAEA.

  12. 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)

  13. Results obtained in reactor safety research for German light-water reactors

    International Nuclear Information System (INIS)

    The author demonstrates the present-day state of light-water-reactor safety research using results obtained in the Kernforschungszentrum Karlsruhe (KfK) throughout the last 20 years. The first topic concerns the field of 'loss-of-coolant accident' which is directly related to the design of light-water reactor. The best part of this paper, however, is dedicated to topics dealing with beyond-design reactor loads which are investigated in risk studies, with special reference to research results obtained for core melt down and to corresponding assessments of the release of radioactive substances to the environment (source terms). The author finally assesses the state of research and outlines discernable trends in light-water-reactor safety research. (orig./HP)

  14. 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

  15. 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.

  16. Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor

    International Nuclear Information System (INIS)

    The paper describes elements of design consideration of supercritical-pressure, light water cooled reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the reactors are 1. Compact reactor and turbines due to high specific enthalpy of supercritical water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water cooled reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast reactor core due to small coolant flow rate, potentially easy shift to fast breeder reactor without changing coolant technology. 10.Potential of producing energy products such as hydrogen and high quality hydro carbons. (authors)

  17. 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...

  18. 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

  19. 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)

  20. Generate light with wind power

    OpenAIRE

    Iqbal, Fowad

    2013-01-01

    The report explain the steps taken to improve a product (SOLVINDEN), which uses sun and wind energy to generate light and is used for outdoor decoration. The research involves improvements in both designas well function. As the form follows function in the product functionality of the form is very important in selection of the form. Some of important topics which are considered are different way of using wind to charge batteries, blades profiles and shape, way of optimizing generator, ratio o...

  1. Hybrid reactor blankets for constant energy multiplication and flat power distribution

    International Nuclear Information System (INIS)

    Two blanket design difficulties are usually attributed to the blanket neutronic properties: high peak-to-average power density ratio distribution and the variation of the energy multiplication with burnup. This work shows that blankets can be designed to have a constant energy multiplication and a flat power distribution. These features are illustrated for light water hydrid reactor blankets

  2. Safety research for evolutionary light water reactors

    International Nuclear Information System (INIS)

    The development of nuclear energy has been characterized by a continuous evolution of the technological and philosophical underpinnings of reactor safety to enable operation of the plant without causing harm to either the plant operators or the public. Currently, the safety of a nuclear plant is assured through the combined use of procedures and engineered safety features together with a system of multiple protective barriers against the release of radioactivity. This approach is embodied in the concept of Design-Basis Accidents (DBA), which requires the designers to demonstrate that all credible accidents have been identified and that all safety equipment and procedures perform their functions extremely reliably. Particularly important functions are the automatic protection to shut the reactor down and to remove the decay heat while ensuring the integrity of the containment structure. Within the DBA concept, the so-called severe accidents were conveniently defined to be those accidents that lie beyond the DBA envelope; hence, they did not form part of the safety case. (author)

  3. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    The reactor concept analyzed is a 233UF6 core surrounded by a molten salt (Li7F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  4. UF6 breeder reactor power plants for electric power generation

    Science.gov (United States)

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

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  5. 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

  6. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

  7. IRIS - Generation IV Advanced Light Water Reactor for Countries with Small and Medium Electricity Grids

    International Nuclear Information System (INIS)

    An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a Generation IV Reactor, International Reactor Innovative and Secure (IRIS). IRIS is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., fuel cycle sustainability, enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it does not require new technology development since it relies on the proven technology of light water reactors. This paper presents the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and four-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. The path forward for possible future extension to a eight-year cycle will be also discussed. IRIS has a large potential worldwide market because of its proven technology, modularity, low financing, compatibility with existing grids and very limited infrastructure requirements. It is especially appealing to developing countries because of ease of operation and because its medium power is more adaptable to smaller grids. (author)

  8. Dual pressurized light water reactor producing 2000 M We

    International Nuclear Information System (INIS)

    The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

  9. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  10. Shedding Light on Solar Power

    Science.gov (United States)

    2002-01-01

    Glenn Research Center sponsored an SBIR contract with ENTECH, in which the company worked to mold its successful terrestrial concentrator technology into applications that would generate solar power for space missions. ENTECH's first application made use of small, dome-shaped Fresnel lenses to direct sunlight onto high- efficiency photovoltaic cells. After some key adjustments, the mini- dome lens array was flown as part of the U.S. Air Force/NASA Photovoltaic Array Space Power Plus Diagnostics (PASP Plus) flight experiment in 1994. Due to their three-dimensional shape, the mini- dome lenses entailed construction by a batch molding process, which is naturally more costly than a continuous process. To overcome this disadvantage and meet the requirement for precise solar pointing in two axes, ENTECH started developing solar concentrator arrays for space using a line-focus lens that can be mass-produced by a continuous process. This new technology, named Solar Concentrator Array with Refractive Linear Element Technology (SCARLET), was created with support from Glenn and the Ballistic Missile Defense Organization, and was used to power the NASA/Jet Propulsion Laboratory Deep Space 1 spacecraft.

  11. 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)

  12. 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)

  13. 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.

  14. 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)

  15. 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.

  16. 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

  17. Pellet bed reactor for multi-modal space power

    International Nuclear Information System (INIS)

    A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in tens of megawatts. It is envisioned that the electrical power requirements will be two-fold; long-duration low power will be needed for station keeping, communications and/or surveillance, while short-duration high power will be required for pulsed power devices. These power characteristics led to authors to propose a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multi-megawatt reactor power source are the following: standby, alert and pulsed power modes; high thermal output heat source (around 1000 MWt peak power); long lifetime standby power (10-30 yrs); high temperature output (1500-1750 K); rapid burst power transition; high reliability (>95%); and meets stringent safety requirements. The proposed pellet bed reactor concept is designed to satisfy these characteristics

  18. 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

  19. 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.

  20. 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

  1. 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

  2. International students workshop on innovative light water reactors

    International Nuclear Information System (INIS)

    Nuclear reactor design is still one of the most fascinating subjects of mechanical engineering. Thirty students from 10 worldwide nations demonstrated this impressively in a recent workshop on supercritical water cooled reactors of the 4th generation, held from March 31 to April 3, 2008, in Karlsruhe, Germany, hosted by the Karlsruhe Institute of Technology. Bachelor and master students as well as young scientists working on their doctorate presented their own particular contribution to design and analyses of innovative reactor components, including its safety systems and other plant design. Their presentations were accompanied by lectures of leading scientists working in the European project of the 'High Performance Light Water Reactor' which is sponsored by the European Commission as part of its 6th Framework Programme. The workshop is an initiative of the Generation IV International Forum. (orig.)

  3. Nuclear problems of power reactors safety

    International Nuclear Information System (INIS)

    The objective of this presentation is to emphasize the conditions that would be of high importance in safety analyses concerned first of all with reactor core. It describes reactor kinetics processes in the core and build up of fission products, classification of reactor accidents related to the core, risk estimation and includes a list of importance reactor accidents

  4. PARs for combustible gas control in advanced light water reactors

    International Nuclear Information System (INIS)

    This paper discusses the progress being made in the United States to introduce passive autocatalytic recombiner (PAR) technology as a cost-effective alternative to electric recombiners for controlling combustible gas produced in postulated accidents in both future Advanced Light Water Reactors (ALWRs) and certain U. S. operating nuclear plants. PARs catalytically recombine hydrogen and oxygen, gradually producing heat and water vapor. They have no moving parts and are self-starting and self-feeding, even under relatively cold and wet containment conditions. Buoyancy of the hot gases they create sets up natural convective flow that promotes mixing of combustible gases in a containment. In a non-inerted ALWR containment, two approaches each employing a combination of PARs and igniters are being considered to control hydrogen in design basis and severe accidents. In pre-inerted ALWRs, PARs alone control radiolytic oxygen produced in either accident type. The paper also discusses regulatory feedback regarding these combustible gas control approaches and describes a test program being conducted by the Electric Power Research Institute (EPRI) and Electricite de France (EdF) to supplement the existing PAR test database with performance data under conditions of interest to U.S. plants. Preliminary findings from the EPRI/EdF PAR model test program are included. Successful completion of this test program and confirmatory tests being sponsored by the U. S. NRC are expected to pave the way for use of PARs in ALWRs and operating plants. (author)

  5. 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.

  6. 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)

  7. 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.)

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

    International Nuclear Information System (INIS)

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

  9. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Topics that have been considered during this year include (1) stress corrosion cracking (SCC) of austenitic stainless steels (SS), (2) fatigue of Type 316NG SS, and (3) SCC of ferritic steels used in reactor piping, pressure vessels, and steam generators. Constant-extension-rate-tests (CERTs) and crack-growth-rate (CGR) tests on 1TCT specimens were performed to quantify the effects of chromate, sulfate, chloride, and low levels of organic impurities on SCC of sensitized Type 304 and Type 316NG SS. Some organic substances were found to actually inhibit SCC of sensitized Type 304 SS. Fatigue tests on Type 316NG SS showed that even at 0.5 Hz the fatigue life in the environment is about half of that in air, and the reduction in life increases at lower frequencies. CGR tests on A533-Gr B pressure vessel steel are in progress on conventional specimens and composite specimens of A533-Gr B/Inconel-182/Inconel-600. Some of the specimens have been plated with either nickel or gold to reduce contact between the surface of low-alloy steel and the environment. Comparison of the results from the bare and plated specimens will provide insight into whether electron transfer through the oxide film on the bulk surface of the ferritic steel is important in the overall SCC process. 49 refs., 16 figs., 9 tabs

  10. 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

  11. 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.

  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. 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

  14. 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

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M. [Nuclear Science Program, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia)

    2015-04-29

    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.

  17. 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.

  18. 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

  19. Requirements for next generation light water reactors for the 21st century in Japan

    International Nuclear Information System (INIS)

    Although nuclear power generation with light water reactors has been successfully employed and developed in Japan, further improvements in economics and safety are desirable in order to achieve further development and maturity of nuclear power. At the present time advanced light water reactors have successfully been developed and construction of two units has started in Japan. Large LWRs of highly developed function and performance have been studied as a candidate for the next-generation power reactor adopting the newest technological concepts as for Advanced LWRs such as ABWR and APWR. On the other hand, innovative design ideas which differ from conventional concepts are being advocated abroad and in Japan. Issues on the energy situation, labor force, etc. likely to arise in the 21st century in relation to the design of LWRs and of nuclear power development and requirements for Japan are discussed. The anticipated social changes toward the future have been considered as the surrounding conditions for defining the required specifications of the new-concept reactors. The results of the discussion are summarized below for the items most likely to be encountered from the aspects of energy demand, labor issues and technical development. (author). 4 figs

  20. Reactor and turbine building layout of the high performance light water reactor

    International Nuclear Information System (INIS)

    Based on the information generated within the European funded project ''High Per-formance Light Water Reactor Phase 2'', a general plant layout has been developed. The central building is the reactor building, in which the containment and safety sys-tems are located. The reactor building is with app. 90.000 m3 considerably smaller compared to other BWR buildings, thus providing a huge potential for cost savings. The turbine building with app 250,000 m3 is of approximately the same size like for existing BWRs. (orig.)

  1. 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

  2. Environmental impact of a suitable nuclear power reactor used to provide a process heat system to synthesize fuels

    International Nuclear Information System (INIS)

    The environmental impact of an 1100-MWE light water nuclear power reactor used to provide a process heat system to synthesize fuels is evaluated covering thermal pollution, visual pollution, and radioactive water quantities produced and released to the environment

  3. Light water reactor safety. Past, present and future

    International Nuclear Information System (INIS)

    This paper presents a review of the past, present and possible future developments in light water reactor (LWR) safety. The paper divides the past into two periods: the distant past i.e., before the TMI-2 accident when the main concern was with the design basis, the general design criteria, the concept of the defense in depth, the thermal hydraulics of the large loss of coolant accident (LOCA) and the success of the emergency core cooling system (ECCS), and the near past, i.e., after the TMI-2 accident when the main concern was with the physics of the postulated severe accidents: their prevention and mitigation. The present period is chosen as the translation of the research on the design basis and severe accidents into practical designs of Gen III+ with their core catchers and severe accident management (SAM) strategies, which could, in fact, provide ample assurances of public safety even for very severe accidents. The paper attempts to describe the remaining safety issues for both the Gen II and Gen III+ nuclear plants. The more important safety challenges are being posed by the recent moves of (1) extension of the life of the presently installed Gen II LWRs to 60 years (and perhaps to 80 years) and (2) the large uprates in power that are being sought for the Gen II LWRs. Clearly, the safety margins will be tested by these moves of long extended operations with greater power ratings of the Gen II plants. A prognosis of the emerging development trends in the LWR safety has been attempted with some suggestions. (author)

  4. 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)

  5. An overview of future sustainable nuclear power reactors

    OpenAIRE

    Andreas Poullikkas

    2013-01-01

    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 ...

  6. Reactor Physics Tests for the Full Power Operation of HANARO

    International Nuclear Information System (INIS)

    The initial criticality of HANARO was achieved on the Feb. 8th of 1995. As HANARO is a unique reactor, there were difficulties to get a license to its full power operation, in which the design power of HANARO is 30 MW. There were two operation license conditions that limited the operation power to 80% of the design power. They were resolved in 2003 and the power ascension tests were conducted for the full power operation. This paper presents the several reactor physics tests for the power ascension to the full power of HANARO

  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. 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

  9. 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.

  10. Medium-Power Lead-Alloy Reactors: Missions for This Reactor Technology

    International Nuclear Information System (INIS)

    A multiyear project at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology investigated the potential of medium-power lead-alloy-cooled technology to perform two missions: (1) the production of low-cost electricity and (2) the burning of actinides from light water reactor (LWR) spent fuel. The goal of achieving a high power level to enhance economic performance simultaneously with adoption of passive decay heat removal and modularity capabilities resulted in designs in the range of 600-800 MW(thermal), which we classify as a medium power level compared to the lower [∼100 MW(thermal)] and higher [2800 MW(thermal)] power ratings of other lead-alloy-cooled designs. The plant design that was developed shows promise of achieving all the Generation-IV goals for future nuclear energy systems: sustainable energy generation, low overnight capital cost, a very low likelihood and degree of core damage during any conceivable accident, and a proliferation-resistant fuel cycle. The reactor and fuel cycle designs that evolved to achieve these missions and goals resulted from study of the following key trade-offs: waste reduction versus reactor safety, waste reduction versus cost, and cost versus proliferation resistance. Secondary trade-offs that were also considered were monolithic versus modular design, active versus passive safety systems, forced versus natural circulation, alternative power conversion cycles, and lead versus lead-bismuth coolant.These studies led to a selection of a common modular design with forced convection cooling, passive decay heat removal, and a supercritical CO2 power cycle for all our reactor concepts. However, the concepts adopt different core designs to optimize the achievement of the two missions. For the low-cost electricity production mission, a design approach based on fueling with low enriched uranium operating without costly reprocessing in a once-through cycle was pursued to achieve a

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

    International Nuclear Information System (INIS)

    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

  12. 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

  13. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    International Nuclear Information System (INIS)

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and 233U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles

  14. 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

  15. Nuclear modular power stations with lead-based coolant reactors

    International Nuclear Information System (INIS)

    In the present report, the projects of reactors with the lead-based coolant are considered. This class of reactors has the advantages and limitations. The main of advantages is enhanced safety and the main of restrictions is the limitation on power, both originate from natural properties of lead-based coolants. This limitation uniquely determines lead-technology reactors as medium - and small-power systems. (author)

  16. Automated procedure for selection of optimal refueling policies for light water reactors

    International Nuclear Information System (INIS)

    An automated procedure determining a minimum cost refueling policy has been developed for light water reactors. The procedure is an extension of the equilibrium core approach previously devised for pressurized water reactors (PWRs). Use of 1 1/2-group theory has improved the accuracy of the nuclear model and eliminated tedious fitting of albedos. A simple heuristic algorithm for locating a good starting policy has materially reduced PWR computing time. Inclusion of void effects and use of the Haling principle for axial flux calculations extended the nuclear model to boiling water reactors (BWRs). A good initial estimate of the refueling policy is obtained by recognizing that a nearly uniform distribution of reactivity provides low-power peaking. The initial estimate is improved upon by interchanging groups of four assemblies and is subsequently refined by interchanging individual assemblies. The method yields very favorable results, is simpler than previously proposed BWR fuel optimization schemes, and retains power cost as the objective function

  17. 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

  18. 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

  19. 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.)

  20. 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

  1. 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.

  2. Probabilistic analysis for the Babcock and Wilcox advanced light water reactor

    International Nuclear Information System (INIS)

    The Babcock and Wilcox (B and W) Advanced Light Water Reactor (ALWR) design employs design features that will provide enhanced safety, reliability, and design margin over the current generation of commercial nuclear power plants. This paper presents a probabilistic analysis performed to provide early feedback to the designers to enhance the reliability of these systems. Feedback from the probabilistic analysis was used to improve the system design by incorporating the insights gained. The calculated core melt frequency for the ALWR design was better than the design targets since most of the features that dominate the risk profile in conventional pressurized water reactors (PWRs) were eliminated in the redesign for the ALWR

  3. 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)

  4. 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.

  5. Radioactive waste produced from Integral Fast Reactor: Comparisons to Light-Water Reactors

    International Nuclear Information System (INIS)

    The main goal of this study was to compare between radioactive waste that produced from Integral Fast Reactor (IFR) and Light-Water Reactors (LWR). The radioactive waste produced from IFR reactors either have a short half life, which means that they decay quickly and become relatively safe, or a long half life, which means that they are only slightly radioactive. The primary argument for pursuing IFR-style technology today is that it provides the best solution to the existing nuclear waste problem because breeder reactors can be fueled from the waste products of existing reactors as well as from the plutonium used in weapons. Depleted uranium (DU) waste can also be used as fuel in IFR reactors. IFR-style reactors produce much less waste than LWR-style reactors, and can even consume other waste as fuel. The total volume of fission products is 1/20th the volume of used fuel produced by a light water plant of the same size, and considered to be waste. 70% of fission products are either stable or have half lives less than one year. Technetium-99 and iodine-129, which constitute 6% of fission products, have very long half lives but can be transmuted to isotopes with very short half lives (15.46 seconds and 12.36 hours) by neutron absorption within a reactor, effectively destroying them. Zirconium-93, another 5% of fission products, could in principle be recycled into fuel-pin cladding, where it doesn't matter that it is radioactive. The remaining high level waste from reprocessing, about 200kg per GWe-yr, is less radiotoxic than mined uranium within 400 years. The radioactivity of the waste decays to levels similar to the original ore in about 200 years. (author)

  6. 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)

  7. 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

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.U.; Layman, W.H. (Electric Power Research Institute, Palo Alto, CA (United States))

    1992-01-01

    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.

  11. 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

  12. The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

    International Nuclear Information System (INIS)

    If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issue through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MWe IFR capacity for every three MWe Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years)

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. 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 ...

  18. High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

    International Nuclear Information System (INIS)

    The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

  19. 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.

  20. 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)

  1. Quality surveillance for PWR power plant reactor internals manufacturing

    International Nuclear Information System (INIS)

    The structure and the function of the reactor internals of the improved generation Ⅱ PWR power plant is instructed briefly, the critical factors and difficulties in the manufacture process for reactor internals are analyzed, the quality control and surveillance of reactor internals manufacturing is discussed, especially the critical factors and difficulties of the quality control in the manufacture process for the main parts of reactor internals and in the reactor internals assembling process are represented in detail, the key points of the resident manufacture supervision is presented, and other key points of quality control in the manufacture process are also given, such as the documents control and personnel control. (author)

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

    International Nuclear Information System (INIS)

    The regulations applying to reactors for power generation mentioned in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. Covered are the following: definitions of terms, application for the permission to install reactors, application for the permission to alter installations, reactor operation plans, keeping of various records, limitations on ingress and ingress in radiation controlled areas, measures concerning radiation exposure doses, operation of reactors, on-site transport, storage of nuclear fuel materials and radioactive waste, security regulations, steps taken during times of danger, making of various reports, and so on. (Mori, K.)

  3. Distribution of characteristics of LWR [light water reactor] spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Reich, W.J.; Notz, K.J. [Oak Ridge National Lab., TN (USA); Moore, R.S. [Automated Sciences Group, Inc., Oak Ridge, TN (USA)

    1991-01-01

    The purpose of this report is to develop a collective description of the entire spent fuel inventory in terms of various fuel properties relevant to Approved Testing Materials (ATMs) using information available from the Characteristics Data Base (CBD), which is sponsored by the US Department of Energy`s (DOE`s) Office of Civilian Radioactive Waste Management. A number of light-water reactor (LWR) characteristics were analyzed including assembly class representation, fuel burnup, enrichment, fuel fabrication data, defective fuel quantities, and, at PNL`s specific request, linear heat generation rate (LHGR) and the utilization of burnable poisons. A quantitative relationships was developed between burnup and enrichment for BWRs and PWRs. The relationship shows that the existing BWR ATM is near the center of the burnup-enrichment distribution, while the four PWR ATMs bracket the center of the burnup range but are on the low side of the enrichment range. Fuel fabrication data are based on vendor specifications for new fuel. Defective fuel distributions were analyzed in terms of assembly class and vendor design. LHGR values were calculated from utility data on burnup and effective full-power days; these calculations incorporate some unavoidable assumptions which may compromise the value of the results. Only a limited amount of data are available on burnable poisons at this time. Based on this distribution study, suggestions for additional ATMs are made. These are based on the class and design concepts and include BWR/2,3 barrier fuel, and the WE 17 {times} 17 class with integral burnable poison. Both should be at relatively high burnups. 16 refs., 5 figs., 15 tabs.

  4. Nuclear Power Reactors in the World. 2012 Ed

    International Nuclear Information System (INIS)

    This is the 32nd edition of Reference Data Series No. 2, which presents the most recent reactor data available to the IAEA. It contains summarized information as of the end of 2011 on power reactors that are in operation, under construction and shut down, and performance data on reactors operating in IAEA Member States, as reported to the IAEA. The information is collected through designated national correspondents in the Member States and the data area used to maintain the IAEA's Power Reactor Information System.

  5. 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)

  6. 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

  7. Adaptation of WWR-K reactor for irradiation work at low power

    International Nuclear Information System (INIS)

    According to INAA in Sept 2000 there were 288 research reactors in operation in the world. 67 % of these reactors were operating in power mode below 1.1 MW. The profile of an available scientific and technological program depends to a great extent on reactor power. High-power reactors are used mainly for isotope production and material testing, specially for needs of power generation programs. Low-power reactors are used mainly for education and for irradiation for purposes of activation analysis. TRIGA, Slowpoke and the Chinese Mini Reactor are typical instruments in this class. High neutron flux can also be utilized for activation analysis. In this case, samples for irradiation should be tightly sealed in aluminium or in quartz, and organic samples can be irradiated for a relatively short time only. Another important issue is that reactor operation on a high power level is connected with higher expenses for fuel, for electric power, for coolant and for general maintenance. Many reactor centers use to charge prices like USD 200 to 400 for 1 day irradiation of a single container 10-12 cm high and 2.5 to 5 cm in diameter. For many users this price is prohibitive. In a low-power reactor all the expenses are smaller and samples of organic origin can be much larger. The WWR-K is 6 MW light water research reactor with the maximum neutron flux 1.4·1014 n·cm-2·s-1 and, correspondingly, rather high operation expenses. However, it can be operated at lower power levels, with much less electric power consumption for cooling systems. To check a possibility of use of the WWR-K as a low-power reactor, to decrease expenses for activation analysis, three experimental runs were performed on 200 k W power. Samples were irradiated in a specially built aluminium container 6 cm in diameter and 60 cm high. Inside the aluminium container there were located 6·120 ml bottles (jars). The total volume available in one irradiation was 720 ml. Analysis of economical and technological

  8. 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.)

  9. Power plant systems for fusion reactors

    International Nuclear Information System (INIS)

    To investigate and compare the applicability of thermal cycles for power generation to nuclear fusion, four plant systems, i.e. direct steam turbine, in-direct steam turbine, direct gas turbine and in-direct steam turbine with gas-cooled blanket, have been designed with the estimates of their thermal efficiencies. An plant designs here are based on the same power core: fusion power 2300 MW, external heating power 58 MW, thermal power of blanket 2420 MW and divertor 490 MW. In addition, it is assumed that the construction of the power plant is near future so that the structural material would be a ferritic/martensitic steel such as F82H to be used at temperatures lower than 500-550degC. Also the divertor is always cooled by water at pressure of 10 MPa, and inlet/outlet temperature of 150-200degC/200-250degC. The removal heat from the divertor is utilized to heat the coolant fed to the blanket inlet in all above plants. The direct steam turbine cycle employs supercritical pressure water at 25 MPa and blanket inlet/outlet temperatures of 280degC/500degC. The steam out from the blanket directly flows into a high pressure turbine. The steam intermediately extracted from the high pressure turbine and/or a part of main steam from the blanket outlet is utilized to reheat the steam coming out of the high pressure turbine. Also the regenerative cycle is applied by using steams extracted from high, medium and low pressure turbines. Eventually the obtained thermal efficiency is 41.4%. The in-direct steam turbine cycle consists of the primary loop which removes the heat from the blanket and the secondary (power generation) loop by using a steam generator at their interface. The primary coolant is supercritical pressure water similar to that of above direct steam cycle, i.e. 25 MPa, 290degC/510degC. The secondary coolant is also water but with the condition of a fast breeder fission reactor, i.e. 16.3 MPa, 210degC/480degC. With reheat and regenerative cycle, the thermal

  10. 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.

  11. Comparative study of plutonium burning in heavy and light water reactors

    International Nuclear Information System (INIS)

    There is interest in the U.S. and world-wide in reducing the burden on geological nuclear fuel disposal sites. In some disposal scenarios, the decay heat loading of the surrounding rock limits the commercial spent fuel capacity of the sites. In the long term (100 to 1,500 years), this decay heat is generated primarily by actinides, particularly 241Am and 241Pu. One possible approach to reducing this decay-heat burden would be to reprocess commercial spent nuclear fuel and use intermediate-tier thermal reactors to 'burn' these actinides and other transuranics (plutonium and higher actinides). The viability of this approach is dependent on the detailed changes in chemical and isotopic compositions of actinide-bearing fuels after irradiation in thermal reactor spectra. The intermediate-tier thermal burners could bridge the commercial water-cooled reactors and fast reactors required for ultimate consumption of the transuranics generated in the commercial reactors. This would reduce the number of such fast reactors required to complete the mission of burning transuranics. If thermal systems are to be used for the transmutation mission, it is likely that they would be similar to or are advanced versions of the systems currently used for power generation. In both the U.S. and Canada, light- and heavy-water-cooled thermal reactors are used for power generation in the commercial nuclear sector. About 103 pressurized- and boiling- light water reactors (PWRs and BRWs) are deployed in the U.S. nuclear industry while about 18 CANDU (heavy-water-cooled) reactors are used in the Canadian industry. There are substantial differences between light and heavy water-cooled reactors that might affect transmutation potential. These arise from differences in neutron balance of the reactors, in neutron energy spectra, in operational approaches (e.g., continuous refueling enhancing fuel burnup), and so on. A systematic study has been conducted to compare the transmutation potentials of

  12. Improvements in the management of safety in research reactor operation through appropriate application of selected power reactor good practices

    International Nuclear Information System (INIS)

    Research reactor managers are increasingly implementing improvements in their management of safety through the application of good practices originally developed as power reactor programs. This paper considers ways to select practices to emulate, effectively incorporate them into a research reactor program and evaluate their contribution to safety. Relative to research reactors, power reactor programs look relatively homogeneous when considering source terms, stored energy, core power density, operating cycles, plant systems and staff sizes. They have potential hazard consequences that require effective safety management programs. Finally, power reactors generate a stream of revenue to fund these programs. The power reactor community has combined their resources with the homogeneity of their challenge to create impressive safety management tools, many of which can be effectively implemented in the research reactor community. However, not all programs can be effectively implemented in all research reactors. number of power reactor programs are analyzed in the paper with consideration of their effective implementation and potential contribution to research reactor. (author)

  13. 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

  14. Role of advanced reactors in further nuclear power development

    International Nuclear Information System (INIS)

    As a part of the national long-term nuclear R and D program launched in 1992, an endeavor has been made in Korea to develop advanced nuclear reactor systems with significantly enhanced safety and economics from those of the current generation nuclear power plants. The advanced PWR nuclear reactor systems under development in Korea include 1300 MWe Korean Next Generation Reactor (KNGR), 330 MWt Integral Type System Integrated Modular Advanced Reactor (SMART) for nuclear cogeneration, and 330 MWe Korea Advanced Liquid Metal Reactor (KALIMER) in addition to the evolutionary enhancement of the 1000 MWt KSNPP (Korea Standard Nuclear Power Plant). Three point design philosophy has been adopted for the development of the advanced reactors in Korea : enhancements on safety, economics and public acceptance of nuclear power. To enhance the safety of the advanced reactor systems, a strategy has been adopted to employ advanced design features as well as the passive safety design features. Economically viable design concepts also have been implemented in the evolutionary KSNPP, KNGR, and the SMART development. Economic competitiveness against the fossil plants also has been set as a major objective of the ALWR development program in Korea. These safer and more economical advanced reactors will better promote the public acceptance of the commercial use of the nuclear power and thus could be utilized to meet the forecasted national energy need in the early 21st century. International cooperation in the areas of ALWR development as well as improving public acceptance of the nuclear power is required. (author)

  15. Data management for spent fuel from power reactors in Argentina

    International Nuclear Information System (INIS)

    Updated data for the spent fuel management from the operating power reactors - as of December 31st, 2004 - as well as research reactors - as of March 1st, 2003 - in Argentina are presented. Data for the power reactor spent fuel are received from the nuclear power plant operator (Nucleoelectrica Argentina S.A.) twice a year for the cumulative spent fuel arising up to June 30th and December 31st. Data for the research reactor spent fuel are collected once a year from CNEA operators. At the time being, such data are not managed in a database management system but some of them are handled with a spreadsheet program in order to get total, average, lower and higher values. These values are being used to built the input of codes for calculating the composition, activity and thermal power for the spent fuel as a whole as well as the mass, activity and thermal power for spent fuel elements or nuclides. (author)

  16. 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.

  17. 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.

  18. 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

  19. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors)

  20. Tokamak power systems studies: A second stability power reactor

    International Nuclear Information System (INIS)

    A number of innovative physics and engineering features have been studied which promise to greatly improve the reactor prospects of tokamaks relative to STARFIRE. A reference design point has been developed with the following features: large aspect ratio (A = 6); high beta (β ≅ 0.20), with only mild shaping and no indentation, which brings the maximum toroidal field down to 7 T; low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and EF coil system; and steady-state operation with combined fast wave and lower hybrid wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with control of the current achieved by the appropriate choice of wave frequencies and spectra. By selecting an axial safety factor q(o) = 2.0, MHD stability has been found above β ≅ 0.20. Additional features include: impurity control with self-pumped limiters which bury helium on continuously deposited metal surfaces; liquid Li-cooled blanket which provides good performance with low pressure operation; vanadium alloy blanket structure for higher thermal efficiency (eta = 0.42), longer lifetime and reduced activation; and reduced reactor mass (higher power density) due to smaller TF coil, less shielding, fewer blanket penetrations, and higher wall loading. At low neutron wall loads this device represents a minimum capital cost unit. However, economies of scale are strong, and eventually higher wall loads (W ≅ 8 MW/m2, P/sub net/ = 1400 MW) may prove most attractive. Preliminary investigations show inherently safe operation is likely at W ≥ 5 MW/m2. 15 refs., 3 figs., 1 tab

  1. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author)

  2. The next generation of power reactors - safety characteristics

    International Nuclear Information System (INIS)

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

  3. EBT reactor characteristics consistent with stability and power balance requirements

    International Nuclear Information System (INIS)

    This paper summarizes the results of a recent EBT reactor study that includes both ring and core plasma properties and consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters

  4. Nuclear power reactors in the world. April 2003 ed

    International Nuclear Information System (INIS)

    This is the twenty third edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 2002 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and available at the Internet address http://www.iaea.or.at/programmes/a2

  5. Nuclear power reactors in the world. April 2001 ed

    International Nuclear Information System (INIS)

    This is the twenty-first edition of Reference Data Series No.2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 2000 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and is available at the Internet address http://www.iaea.or.at/programmes/a2

  6. Nuclear power reactors in the world. April 2002 ed

    International Nuclear Information System (INIS)

    This is the twenty-second edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 2001 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and available at the Internet address http://www. iaea.or.at/programmes/a2

  7. 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)

  8. Power coefficient of reactivity in CANDU 6 Reactors

    International Nuclear Information System (INIS)

    The Power Coefficient of Reactivity (PCR) measures the change in reactor core reactivity per unit change in reactor power and is an integral quantity which captures the contributions of the fuel temperature, coolant void and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between the inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity in all operating conditions are maintained within a safe range. The CANDU reactor design takes advantage of the inherent nuclear characteristics of small reactivity coefficient, minimal excess reactivity and very long prompt neutron lifetime to mitigate the magnitude of the demand on the engineered systems for controlling reactivity. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with its design characteristics, such that the overall design of the reactor does not depend on the sign of the PCR. This is a contrast to other reactor design concepts which are dependent on a PCR which is both large and negative in the design of their engineered systems for controlling reactivity. It will be demonstrated that during a Loss of Regulation Control (LORC) event, the impact of having a positive power coefficient, or of hypothesizing a PCR larger than that estimated for CANDU, has no significant impact on the reactor safety. Since the CANDU 6 PCR is small, its role in the operation or safety of the reactor is not significant

  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. 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

  11. Design for reactor core safety in nuclear power plants

    International Nuclear Information System (INIS)

    This Guide covers the neutronic, thermal, hydraulic, mechanical, chemical and irradiation considerations important to the safe design of a nuclear reactor core. The Guide applies to the types of thermal neutron reactor power plants that are now in common use and fuelled with oxide fuels: advanced gas cooled reactor (AGR), boiling water reactor (BWR), pressurized heavy water reactor (PHWR) (pressure tube and pressure vessel type) and pressurized water reactor (PWR). It deals with the individual components and systems that make up the core and associated equipment and with design provisions for the safe operation of the core and safe handling of the fuel and other core components. The Guide discusses the reactor vessel internals and the reactivity control and shutdown devices mounted on the vessel. Possible effects on requirements for the reactor coolant, the reactor coolant system and its pressure boundary (including the pressure vessel) are considered only as far as necessary to clarify the interface with the Safety Guide on Reactor Coolant and Associated Systems in Nuclear Power Plants (IAEA Safety Series No. 50-SG-D13) and other Guides. In relation to instrumentation and control systems the guidance is mainly limited to functional requirements

  12. 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

  13. 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.

  14. Ultrasonic level and temperature sensor for power reactor applications

    International Nuclear Information System (INIS)

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel

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

    International Nuclear Information System (INIS)

    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 UF6 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 233U born from thorium. Fission product removal was continuous. Newly born 233U 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 233U. 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

  16. 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...

  17. The JLab high power ERL light source

    Science.gov (United States)

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

    2006-02-01

    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 ˜ 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 [Carr, et al., Nature 420 (2002) 153]. 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 [Neil, et al., Phys. Rev. Lett. 84 (2000) 662]: up to 10 kW of average power in the IR from 1 to 14 μm 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 ms 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

  18. 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

  19. 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)

  20. Radionuclides in United States commercial nuclear power reactors

    International Nuclear Information System (INIS)

    In the next ten to twenty years, many of the commercial nuclear power reactors in the United States will be reaching their projected lifetime of forty years. As these power plants are decommissioned, it seems prudent to consider the recycling of structural materials such as stainless steel. Some of these materials and components have become radioactive through either nuclear activation of the elements within the components or surface contamination with radioactivity form the operational activities. In order to understand the problems associated with recycling stainless steel from decommissioned nuclear power reactors, it is necessary to have information on the radionuclides expected on or in the contaminated materials. A study has been conducted of radionuclide contamination information that is available for commercial nuclear power reactors in the United States. There are two types of nuclear power reactors in commercial use in the United States, pressurized water reactors (PWRs) and boiling water reactors (BWRs). Before presenting radionuclide activities information, a brief discussion is given on the major components and operational differences for the PWRs and BWRs. Radionuclide contamination information is presented from 11 PWRs and over 8 BWRs. These data include both the radionuclides within the circulating reactor coolant water as well as radionuclide contamination on and within component parts

  1. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    International Nuclear Information System (INIS)

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  2. Non-linear analysis in Light Water Reactor design

    International Nuclear Information System (INIS)

    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

  3. Evolution of service to the power reactor community at the UFTR - beyond the bottom line

    International Nuclear Information System (INIS)

    The University of Florida Training Reactor (UFTR) is a 100-kW light-water-cooled, graphite-and-light-water-moderated modified argonaut-type reactor. Although currently fueled with highly enriched (93% 235U) fuel, analyses are nearly completed for conversion to low-enriched fuel. The core design consists of individual fuel boxes in each of which flow is constrained which is a relatively unique feature among existing nonpower reactors (NPRs), making measurement of hot channel factors and other core parameters a straightforward exercise for operator trainees. Since first licensed to operate at 10 kill in 1959, this NPR facility has had an active but evolving record of continuous service to the power reactor community. Because of its midsize power level, available flux levels limit utilization for radiation damage and other studies in which utilities have been served by other NPR facilities. As a result, UFTR service to the power reactor community has been concentrated in areas where the peak thermal flux (∼2 x 1012 n/cm2 ·s) at full power does not present a significant limitation

  4. Cermet-fueled reactors for multimegawatt space power applications

    International Nuclear Information System (INIS)

    The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

  5. 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.

  6. Assessment of tritium breeding requirements for fusion power reactors

    International Nuclear Information System (INIS)

    This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios

  7. Evolutionary/advanced light water reactor data report

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. Studies on transferring the safety features of the module reactor to a large power reactor

    International Nuclear Information System (INIS)

    The German industries and research institutions have developed the HTR module reactor, which is strongly characterized by inherent safety features. The power output is limited to about 200 MWth because of its core configuration. It has been investigated in this work, whether the safety features of the module reactor can be transferred to larger power reactors. For this purpose the conceptual design of a ring core pebble bed reactor has been made with a thermal power output of 3000 MW. By means of computer calculations, the principal physical, thermohydraulical and safety features of the ring reactor have been studied. It has been shown that the 3000-MWth ring reactor basically possesses the same safety characteristics as the small module reactor. At reactivity disturbances, the reactor is shut down passively by the strongly negative temperature coefficient. The decay heat removal is also realized based on the passive priniciple. In the case of a total loss of coolant, the maximum fuel element temperature remains below 1600deg C; and consequently the retention of fission products in the fuel elements is fully attained. The control of xenon oscillations takes place inherently due to the mutual coupling between the local power production and the fuel temperature. (orig.)

  10. Preliminary nuclear power reactor technology qualitative assessment for Malaysia

    International Nuclear Information System (INIS)

    Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

  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. 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

  13. 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

  14. 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.).

  15. 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)

  16. 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

  17. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  18. 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

  19. Control and safety of light water reactors fuelled with plutonium

    International Nuclear Information System (INIS)

    The report summarizes the main results of safety analyses which allow to compare the behaviour of plutonium and uranium fuelled cores with particular reference to BWR cores (Garigliano and Caorso) and to a PWR core - Trinovercellese. The control and safety study is focused on the following accidental events: control rod ejection and steam line break for the PWR and control rod drop, failure of the main heat sink heat-up phase of a Loca for the BWRs. Moreover, the report contains the main conclusions of a propaedentic study of the impact of plutonium recycle on light water reactors core safety

  20. 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

  1. The economics of the fuel cycle (light water reactors)

    International Nuclear Information System (INIS)

    The economical characteristics of the fuel cycle (of light water reactors) as well as the definition and calculation method for the average updated cost of the kWh are recalled. The evolution followed by the unit prices of the different operations of the cycle, their total cost and the part taken by this cost in the overall cost of nuclear kWh are described. The effects on the cost of fuel of certain hypotheses, operating requirements and additional cost factors are considered

  2. 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

  3. 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

  4. 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

  5. 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.)

  6. Nuclear power reactor development in Japan

    International Nuclear Information System (INIS)

    Based on the current situation in Japan that nearly 40 LWRs (PWR/BWR nearly 50/50) have been operating with excellent performance; the number of un-intended shutdown is the smallest, leaker fuel is practically zero and thus coolant activity is the lowest in the world, they proceed to introduce advanced LWR (ALWR) of 1,300 MWe in early '90s, followed by the further improvement into higher duty LWR (so-called next generation LWR). National program for advanced thermal reactor (ATR) and FBR development has been steadily performed; a demonstration reactor of ATR and a prototype of FBR are under construction, which are to be put into operation by middle 1990s. Lately-revised high temperature gas cooled reactor program for process heat generation is put into practice in 1989 and its experimental reactor to be used also for irradiation facility will be completed 7 years later. Smaller-size LWR and so-called 'inherent safety type reactor' have also been interested. Overview on those activities are presented. (author)

  7. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  8. Probing light sterile neutrinos in medium baseline reactor experiments

    CERN Document Server

    Esmaili, Arman; Peres, O L G; Tabrizi, Zahra

    2013-01-01

    Medium baseline reactor experiments (Double Chooz, Daya Bay and RENO) provide a unique opportunity to test the presence of light sterile neutrinos. We analyze the data of these experiments in the search of sterile neutrinos and also test the robustness of theta_13 determination in the presence of sterile neutrinos. We show that existence of a light sterile neutrino state improves the fit to these data moderately. We also show that the measured value of theta_{13} by these experiments is reliable even in the presence of sterile neutrinos, and the reliability owes significantly to the Daya Bay and RENO data. From the combined analysis of the data of these experiments we constrain the mixing of a sterile neutrino with mass squared difference in the range of (10^{-3}-10^{-1}) eV^2 to sin^2 2 theta_{14} <0.1 at 95 % C.L..

  9. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  10. Training simulator for nuclear power plant reactor monitoring

    International Nuclear Information System (INIS)

    A description is given of a method and apparatus for the real-time dynamic simulation of a nuclear power plant that includes a control and nuclear instrumentation console for operating the reactor and monitoring three-dimensional physical values in the reactor core. A digital computer is connected to the console to calculate physical values such as nuclear flux, power, and temperature including the distribution thereof throughout the core with such calculations including the effect of full length, part length, and malfunctioned reactor control rods, as well as xenon, decay heat and boron, for example, on the output and distribution of power within the core. The simulation also includes instrumentation that responds to the calculated physical values by recording a continuous trace of the flux value in the reactor core from the top to the bottom

  11. 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

  12. 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

  13. 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.

  14. 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

  15. Aging management of major light water reactor components

    International Nuclear Information System (INIS)

    Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

  16. Application of reactor-pumped lasers to power beaming

    Science.gov (United States)

    Repetti, T. E.

    1991-10-01

    Power beaming is the concept of centralized power generation and distribution to remote users via energy beams such as microwaves or laser beams. The power beaming community is presently performing technical evaluations of available lasers as part of the design process for developing terrestrial and space-based power beaming systems. This report describes the suitability of employing a nuclear reactor-pumped laser in a power beaming system. Although there are several technical issues to be resolved, the power beaming community currently believes that the AlGaAs solid-state laser is the primary candidate for power beaming because that laser meets the many design criteria for such a system and integrates well with the GaAs photodiode receiver array. After reviewing the history and physics of reactor-pumped lasers, the advantages of these lasers for power beaming are discussed, along with several technical issues which are currently facing reactor-pumped laser research. The overriding conclusion is that reactor-pumped laser technology is not presently developed to the point of being technically or economically competitive with more mature solid-state technologies for application to power beaming.

  17. Hiberarchy of requirement analysis of reactor protection system for advanced pressurized water reactor nuclear power plant

    International Nuclear Information System (INIS)

    In order to improve the security and the margin of safety of nuclear power plant, the research on requirement analysis of digital reactor protection system for advanced pressurized water reactor nuclear power plant was developed. Based on the known technology, a requirement analysis report was performed. A kind of three-levels pyramidal hierarchy was adopted in the requirement analysis, and the design characteristics of the requirement analysis were described in the analysis report. This hiberarchy can directly illuminate the design characters and logical achievement of the requirement analysis for advanced pressurized water reactor digital protection system. (authors)

  18. 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

  19. 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

  20. Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor

    OpenAIRE

    2014-01-01

    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 us...

  1. Advances in environmental fatigue evaluation for light water reactor components

    International Nuclear Information System (INIS)

    The recently acquired fatigue test data have indicated that the low cycle fatigue lives attained in light water reactor environments can be markedly shorter than those determined in the ambient atmospheric air. Of the various factors that are known to exert influence on the fatigue life, the present authors have noted strain rate, temperature, and dissolved oxygen (DO) content, and proposed a method of evaluating the environmental fatigue lives for the Class 1 vessels when they vary with time. In this paper, this method is examined against, and revised in view of, the fatigue data acquired since then, and subsequently simplified so as to become adaptable to the Class 1 piping besides the Class 1 vessels. Then, these two versions of this evaluation method are combined into a methodological system by their respective nature, and, in doing so, have proved their worth by applying themselves to several different sorts of component. The results show that the effects of the light water reactor environments on the fatigue life is a factor of 2 or so in terms of the increase of the usage factor (i.e., the environmental effect correction factor Fen = ca. 2)

  2. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  3. Qualification of the ARROTTA code for light water reactor accident analysis

    International Nuclear Information System (INIS)

    Qualification efforts have been performed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research (INER) for the three-dimensional spatial kinetics code ARROTTA for light water reactor (LWR) core transient analysis. Together TPC and INER started a 5-yr project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multidimensional kinetics calculations such as rod ejection for pressurized water reactors and rod drop for boiling water reactors (BWR). To qualify ARROTTA for evaluation of the Final Safety Analysis Report licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicate that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include the Nuclear Energy Agency Committee on Reactor Physics rod ejection problem, the three-dimensional Langenbuch-Maurer-Werner LWR rod withdrawal/insertion problem, and the three-dimensional linear regression analysis BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared with other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multidimensional core transients for LWRs

  4. 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)

  5. 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

  6. 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…

  7. 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

  8. 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)

  9. The core design of the advanced power reactor plus (APR+)

    International Nuclear Information System (INIS)

    Advance Power Reactor Plus (APR+), a pressurized water reactor and an improved nuclear power reactor based on the Advanced Power Reactor 1400 MWe (APR1400) in Korea, has been developed with 18-month cycle operation strategy from its initial core. The APR+ core power is 4290 MWth which corresponds to a 1500 MWe class nuclear power plant. The reactor core consists of 257 fuel assemblies. Comparing with APR1400 core design, 16 fuel assemblies are added. Its cycle length is expected about 450 EFPD directly from initial core, although most of previous other plants had been started according to their annual or 15-month cycle operation schedule at their initial core and gone to 18-month after third - fourth cycle. In order to reduce the peaking power, fuel pin configurations of the assembly, are optimized by using some low enriched fuel pins and gadolinia bearings. APR+ core has been met the requirements as well as the above cycle length requirement; 1) peaking factor, 2) Negative MTC(Moderator Temperature Coefficient), 3) sufficient shutdown margin, 4) convergent Xenon stability Index. The maximum rod burnup and the discharge fuel assembly burnup are also satisfied those of the limit. It is expected to acquire the standard design approval by the end of 2012 by the Korean nuclear regulatory. (authors)

  10. 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 behaviour 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

  11. 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

  12. PC version of PRIS (Power Reactor Information System)

    International Nuclear Information System (INIS)

    The IAEA has been collecting operating experience data on nuclear power plants in the Member States since 1970. In 1980 a computerized database was established, the IAEA Power Reactor Information System (PRIS). To make PRIS data available to the Member States in a more convenient format, the development of a PC version of PRIS started in 1989

  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. 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.

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

    International Nuclear Information System (INIS)

    Reactors based on direct cooled 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 long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) 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. 12 figs

  17. 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.

  18. 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

  19. 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.)

  20. 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.