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Sample records for german boiling-water nuclear

  1. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  2. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    Science.gov (United States)

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  3. Leukemia in the proximity of a German boiling water nuclear reactor: Evidence of population exposure by chromosome studies and environmental radioactivity

    International Nuclear Information System (INIS)

    Dannheim, B.; Heimers, A.; Oberheitmann, B.; Schmitz-Feuerhake, I.; Schroeder, H.; Ziggel, H.

    1997-01-01

    The detection of an exceptional elevation of leukemia in children appearing 5 years after the start-up of the nuclear power plant Kruemmel in 1983, accompanied by a significant increase of leukemia cases in adults gave rise for investigations of radiation exposures of the population living near to the plant. The rate of dicentric chromosomes in peripheral blood lymphocytes of 7 parents of leukemia children and 14 other inhabitants in the proximity of the plant was significantly elevated and showed ongoing exposures over the years of operation. This finding gives rise to the hypothesis that chronic leakages by the reactor had occurred. This assumption is supported by the identification of artificial radioactivity in air, rain water, soil, and vegetation registered by the regular environmental monitoring programme of the nuclear power plant. Calculations of the corresponding source terms show that the originating emissions must have been well above authorized annual limits. The bone marrow dose is supposed to be originated mainly by incorporating of bone-seeking β- and α-emitters. (author)

  4. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Belblidia, L.A.; Chevrier, A.

    1994-01-01

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  5. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

  6. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  8. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    International Nuclear Information System (INIS)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments

  9. Crack formation in ferritic screws of main steam isolation valves in German boiling water reactors

    International Nuclear Information System (INIS)

    Steinmill, H.

    1992-01-01

    In connection with crack formations at screws of main steam isolation valves in boiling water reactors, detected for the first time in 1988 in the Federal Republic of Germany, metallographic and fractographic investigations and coating analyses of screw surfaces and crack flanks were performed in order to find out the causes. These and other investigations of damaged screws were accompanied in the years 1989 and 1990 by autoclave tests made in several laboratories. With a view to the mechanical stress of the screws, tightening tests and stress analyses were performed by means of FEM. Repeated autoclave tests were concluded recently by the Stuttgart MPA. Although these tests are not reported here, it can be stated that the results obtained fit in with the overall framework of the results summed up in this report. With regard to the kind of sample stress and the results obtained, two cases have to be distinguished in the autoclave tests discussed in this report. (orig.) [de

  10. An Order Coding Genetic Algorithm to Optimize Fuel Reloads in a Nuclear Boiling Water Reactor

    International Nuclear Information System (INIS)

    Ortiz, Juan Jose; Requena, Ignacio

    2004-01-01

    A genetic algorithm is used to optimize the nuclear fuel reload for a boiling water reactor, and an order coding is proposed for the chromosomes and appropriate crossover and mutation operators. The fitness function was designed so that the genetic algorithm creates fuel reloads that, on one hand, satisfy the constrictions for the radial power peaking factor, the minimum critical power ratio, and the maximum linear heat generation rate while optimizing the effective multiplication factor at the beginning and end of the cycle. To find the values of these variables, a neural network trained with the behavior of a reactor simulator was used to predict them. The computation time is therefore greatly decreased in the search process. We validated this method with data from five cycles of the Laguna Verde Nuclear Power Plant in Mexico

  11. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    International Nuclear Information System (INIS)

    1986-09-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

  12. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    International Nuclear Information System (INIS)

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner's Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section

  13. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1; Dekontamination des RDB inkl. der Einbauten wie Dampftrockner und Wasserabscheider sowie der angeschlossenen Hilfssysteme im deutschen Siedewasserreaktor ISAR 1

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Michael; Sempere Belda, Luis; Basu, Ashim; Topf, Christian [AREVA GmbH, Erlangen (Germany). Abt. Chemistry Services; Erbacher, Thomas; Hiermer, Thomas; Schnurr, Bernhard; Appeldorn, Thomas van [E.ON Kernkraft GmbH, Kernkraftwerk ISAR, Essenbach (Germany). Abt. Maschinentechnik; Volkmann, Christian [ESG Engineering Services GmbH, Greifswald (Germany)

    2015-12-15

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17{sup th}, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  14. Analysis of water hammer in control rod drive systems of boiling water reactor nuclear power plants

    International Nuclear Information System (INIS)

    Safwat, H.H.; Arastu, A.H.; Lau, S.

    1983-01-01

    The method of characteristics is applied to analyze water hammer in BWR (Boiling Water Reactor) Control Rod Drive (CRD) Systems following fast opening of scram valves. The modelling of the CRD mechanism is presented. Numerical predictions are compared to experimental data. (author)

  15. Two-phase flow in the upper plenum of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    Tinoco, Hernan

    2003-01-01

    The end part of the Emergency Core Spray System (ECSS) of the Boiling Water Reactors (BWRs) at Forsmark Nuclear Power Plant (NPP) is situated in the Upper Plenum. It consists of a pipe network equipped with water injection nozzles. In case of Lost-of-Coolant Accidents (LOCAs), the ECSS should maintain the core covered by water and, at the same time, rapidly cool and decompress the reactor by means of cold water injection. In similar reactors, some welds belonging to the ECSS support have, after a period of time, shown crack indications. Inspection, repair or replacement of these welds is time consuming and expensive. For this reason, it has now been decided to permanently remove the end part of the ECSS and to replace it by water injection in the Downcomer. However, this removal should not be accompanied by undesirable effects like an increase in the moisture of the steam used for operating the turbines. To investigate the effect of this removal on the steam moisture, a CFD analysis of the two-phase flow in the Upper Plenum of Unit 3, with and without ECSS, has been carried out by means of a two-phase Euler model in FLUENT 6.0. The inlet conditions are given by an analysis of the core kinetics and thermal hydraulics by mean of the POLCA-code. The outlet conditions, i. e. the steam separator pressure drops, are given by empirical correlations from the experiments carried out at the SNORRE facility. The predicted the mass flow-rates to each separator, together with empirical correlations for the moisture content of the steam leaving the separators and the steam dryer, indicate a slight decrease in the steam moisture when the ECSS is removed. Also, a minor decrease in pressure losses over the Upper Plenum is achieved with this removal. On the other hand, rounding the sharp edges of the inlet openings to the steam separators at the shroud cover may give a large reduction in pressure losses

  16. Nuclear power plant with boiling water reactor VK-300 for district heating and electricity supply

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Lisitza, F.D.; Romenkov, A.A.; Tokarev, Y.I.

    1998-01-01

    The paper considers specific design features of a pressure vessel boiling water reactor with coolant natural circulation and three-step in-vessel steam separation (at draught tube outlet of the upcomer, within zone of overflow from the upcomer to downcomer and in cyclon-type separators). Design description and analytical study results are presented for the passive core cooling system in the case of loss of preferred power and rupture in primary circuit pipeline. Specific features of a primary containment (safeguard vessel) are given for an underground NPP sited in a rock ground. (author)

  17. Limiting factor analysis of high availability nuclear plants (boiling water reactors). Final report

    International Nuclear Information System (INIS)

    Frederick, L.G.; Brady, R.M.; Shor, S.W.W.; McCusker, J.T.; Alden, W.M.; Kovacs, S.

    1979-08-01

    The pertinent results are presented of a 16-month study conducted for Electric Power Research Institute by General Electric Company, Bechtel Power Corporation, and Philadelphia Electric Company. The study centered around the Peach Bottom 2 Atomic Power Station, but also included limited study of operations at 20 additional operating boiling water reactors. The purpose of the study was to identify and evaluate key factors limiting plant availability, and to identify potential improvements for eliminating or alleviating those limitations. The key limiting factors were found to be refueling activities; activities related to the reactor fuel; reactor scrams; activities related to 20 operating systems or major components; delays due to radiation, turbid water during refueling operations, facilities/working conditions, and dirt/foreign material; and general maintenance/repair of valves and piping. Existing programs to reduce the effect on plant unavailability are identified, and suggestions for further action are made

  18. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1976

    International Nuclear Information System (INIS)

    Scott, R.L.; Gallaher, R.B.

    1977-01-01

    This bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1976. The report includes 1,253 abstracts that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Three of the unique events that occurred during the year are reviewed in detail

  19. Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980

    Energy Technology Data Exchange (ETDEWEB)

    McCormack, K.E.; Gallaher, R.B.

    1982-03-01

    This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

  20. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1976

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.L.; Gallaher, R.B.

    1977-08-02

    This bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1976. The report includes 1,253 abstracts that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Three of the unique events that occurred during the year are reviewed in detail.

  1. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1975

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.L.; Gallaher, R.B.

    1976-07-01

    The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1975. The report includes 1169 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Seven of the unique events that occurred during the year are reviewed in detail.

  2. Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1975

    International Nuclear Information System (INIS)

    Scott, R.L.; Gallaher, R.B.

    1976-07-01

    The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1975. The report includes 1169 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Seven of the unique events that occurred during the year are reviewed in detail

  3. Real-time stability monitoring method for boiling water reactor nuclear power plants

    International Nuclear Information System (INIS)

    Fukunishi, K.; Suzuki, S.

    1987-01-01

    A method for real-time stability monitoring is developed for supervising the steady-state operation of a boiling water reactor core. The decay ratio of the reactor power fluctuation is determined by measuring only the output neutron noise. The concept of an inverse system is introduced to identify the dynamic characteristics of the reactor core. The adoption of an adaptive digital filter is useful in real-time identification. A feasibility test that used measured output noise as an indication of reactor power suggests that this method is useful in a real-time stability monitoring system. Using this method, the tedious and difficult work for modeling reactor core dynamics can be reduced. The method employs a simple algorithm that eliminates the need for stochastic computation, thus making the method suitable for real-time computation with a simple microprocessor. In addition, there is no need to disturb the reactor core during operation. Real-time stability monitoring using the proposed algorithm may allow operation under less stable margins

  4. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  5. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  6. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

    International Nuclear Information System (INIS)

    March-Leuba, J.; Rey, J.M.

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix

  7. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

  8. Safety problems of nuclear power plants with channel-type graphite boiling water reactors

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

    1977-01-01

    Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

  9. Current status and development of heat exchangers for boiling water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Uchiyama, Yoshio; Nishioka, Shuji; Ito, Shizuo

    1975-01-01

    More efficient and reliable operation is required for BWR heat exchangers because of nuclear plant safety and other reasons. Heat exchangers are classified into two categories of systems, the system for normal operation and the system for emergency operation. The present state and future improvement of heat exchangers are described in view of heat transfer performance, material selection, structural design, vibration, and so on. When noncondensing gas exists in vapour, heat transfer performance deteriorates, so that the heat transfer characteristics should be corrected by the adaption of venting the non condensing gas from the system. Heat transfer tubes should have high corrosion resistance to working fluid as well as high thermal conductivity, strength and economy. From that point of view, 30% cupro-nickel tubes will be replaced with 10% cupro-nickel tubes or titanium tubes though some technical development is necessary. These heat exchangers are now designed according to the MITI criteria for the technology concerning nuclear and thermal power generation, ASME Boiler and Pressure Vessel Code Sec. III and some other criteria. Most of heat transfer tube failures are caused from the tube vibration induced by working fluid flow, so that the vibration test and analysis were performed on U-tube elements. Some correction was obtained for design and fabrication techniques. (Iwase, T.)

  10. Model engineering for piping layout of boiling water reactor nuclear station

    International Nuclear Information System (INIS)

    Tsukada, Koji; Uchiyama, Masayuki; Wada, Takanao; Jibu, Noboru.

    1977-01-01

    A nuclear power station is made up of a wide variety of equipment, piping, ventilation ducts, conduits, and cable trays, etc. Even if equipment arrangement and piping layout are carefully planned on drawings, troubles such as interference often occur at field installation. Accordingly, it is thought very useful to make thorough examinations with plastic three-dimensional models in addition to drawings in reducing troubles at field, shortening the construction period, and improving economics. Examination with plastic models offers the following features: (1) It permits visual three-dimensional examination. (2) Group thinking and examination is possible. (3) Troubles due to failure to understand complicated drawings can be reduced drastically. Manufacturing a 1/20 scale model of the reactor building of the Tokai No. 2 Power Station of the Japan Atomic Power Co., Hitachi has performed model engineering-solution of interference troubles related to equipment and piping, securing of work space for in-service inspection (ISI), carry-in/installation of various equipment and piping, and determination of the piping route of which only the starting and terminating points were given under the complicated ambient conditions. Success with this procedure has confirmed that model engineering is an effective technique for future plant engineering. (auth.)

  11. Assessment of water hammer effects on boiling water nuclear reactor core dynamics

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2007-01-01

    Full Text Available Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

  12. Topics to be covered in safety analysis reports for nuclear power plants with pressurized water reactors or boiling water reactors in the F.R.G

    International Nuclear Information System (INIS)

    Kohler, H.A.G.

    1977-01-01

    This manual aims at defining the standards to be used in Safety Analysis Reports for Nuclear Power Plants with Pressurized Water Reactors or Boiling Water Reactors in the Federal Republic of Germany. The topics to be covered are: Information about the site (geographic situation, settlement, industrial and military facilities, transport and communications, meteorological conditions, geological, hydrological and seismic conditions, radiological background), description of the power plant (building structures, safety vessel, reactor core, cooling system, ventilation systems, steam power plant, electrical facilities, systems for measurement and control), indication of operation (commissioning, operation, safety measures, radiation monitoring, organization), incident analysis (reactivity incidents, loss-of-coolant incidents, external impacts). (HP) [de

  13. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, M.P. [Instituto de Estudos Avancados - CTA, Sao Paolo (Brazil); Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.

  14. Development of NUFREQ-N, an analytical model for the stability analysis of nuclear coupled density-wave oscillations in boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Park, G.C.

    1983-01-01

    A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). The model accounts for phasic slip, distributed spacers, subcooled boiling, space/time-dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation. In its final form, this model constitutes a multi-input, multi-output (MIMO) linear system, which features a general nodal neutron kinetics model. Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions. The stability characteristics of a typical BWR/4 has been investigated with the Nyquist criterion. The computer implementation of this mode, NUFREQ-N, was used for the parametric study of a typical BWR/4 and comparison were made with existing in-core and out-of-core data. Also, NUFREQ-N was used to analyze the expected stability characteristics of a typical BWR/4. The parametric results revealed important factors influencing BWR stability margin. It was found that NUFREQ-N generally agreed well with out-of-core data. This was especially true for the predicted power-to-flow transfer function, which is the most important transfer function in thermal-hydraulic stability analysis

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

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

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

  16. Blow-off device for limiting excess pressure in nuclear power plants, especially in boiling water nuclear power plants

    International Nuclear Information System (INIS)

    Simon, U.; Werner, K.D.; Hoffmann, D.; Pontani, B.

    1979-01-01

    In a blow-off device for limiting excess pressure in nuclear power plants, at least one condensation tube disposed so that a lower outflow and thereof is immersed in a volume of water, and an upper inflow end of the condensation tube extends out of the volume of water and is connectible to a source of steam that is to be condensed or a steam-air mixture, the outflow end of the condensation tube, for stabilizing the condensation being provided with an assembly of wall parts forming passageways extending in axial direction for subdividing the steam flow and bubbles produced in the volume of water, the passageways of the assembly of wall parts being stepped in axial direction at both axial ends of the assembly of wall parts

  17. Blow-off device for limiting excess pressure in nuclear power plants, especially in boiling-water nuclear power plants

    International Nuclear Information System (INIS)

    Kuehnel, R.

    1979-01-01

    In a blow-off device for limiting excess pressure in nuclear power plants, at least one condensation tube disposed so that a lower outlet end thereof is immersed in a volume of water in a condensation chamber having a gas cushion located in a space above the volume of water, and the upper inlet end of the condensation tube extending out of the volume of water and being connectible to a source of steam that is to be condensed or a steam-air mixture, the outlet end of the condensation tube, for smoothing the condensation, being provided with wall parts forming passages extending in axial direction, delimited from one another and terminating in the water volume, the wall parts serving to subdivide steam flow from the source thereof and bubbles produced thereby in the water volume, the wall parts being constructed as a tube attachment and being formed with an opening corresponding to the outlet end of the condensation tube and by means of which the tube attachment is mounted on the outlet end of the condensation tube, a first group of the wall parts in the tube attachment being disposed in alignment with the outlet end of the condensation tube, and a second group of the wall parts surrounding the first group thereof, the passages formed by the second group of the wall parts communicating laterally with the passages formed by the first group of the wall parts, the passages formed by the second group of the wall parts, at least at the upper ends thereof, communicating with the water volume

  18. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1990-12-01

    This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs

  19. Boiling water reactor fuel bundle

    International Nuclear Information System (INIS)

    Weitzberg, A.

    1986-01-01

    A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods

  20. Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Conlon, W.M.; Lahey, R.T. Jr.

    1982-12-01

    This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

  1. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  2. Development of BWR [boiling water reactor] and PWR [pressurized water reactor] event descriptions for nuclear facility simulator training

    International Nuclear Information System (INIS)

    Carter, R.J.; Bovell, C.R.

    1987-01-01

    A number of tools that can aid nuclear facility training developers in designing realistic simulator scenarios have been developed. This paper describes each of the tools, i.e., event lists, events-by-competencies matrices, and event descriptions, and illustrates how the tools can be used to construct scenarios

  3. The Swr 1000: a nuclear power plant concept with boiling water reactor for maximum safety and economy of operation

    International Nuclear Information System (INIS)

    Brettschuh, W.

    2001-01-01

    The SWR 1000 is a design concept for a light water reactor nuclear power plant that meets all requirements regarding plant safety, economic efficiency and environ-mental friendliness. As a result of the plant's safety concept, the occurrence of core damage can, for all practical intents and purposes, be ruled out. If a core melt accident should nevertheless occur, the molten core can be retained inside the RPV, thus ensuring that all consequences of such an accident remain restricted to the plant itself. The power generating costs of the SWR 1000 are lower than with those of coal-fired and combined-cycle power plants. Power generation using nuclear energy does not release carbon dioxide to the environment, thus meeting the need for sustainable protection of our global climate. (author)

  4. A model of Altio Lazio boiling water reactor using the LEGO code nuclear steam supply system simulation

    International Nuclear Information System (INIS)

    Garbossa, G.B.; Spelta, S.; Cori, R.; Mosca, R.; Cento, P.

    1989-01-01

    An extensive effort has been made at the Italian National Electricity Board (ENEL) to construct and validate a LEGO model capable of simulating the operational transients of the Alto Lazio Nuclear Station, a two-twin units site with BWR/6 class reactors, rated at 2894 MWt and with Mark III containment. The desired end-product of this effort is an overall plant model consisting of the Nuclear Steam Supply System model, described in this paper, and the Balance of Plant model, capable of simulating the transient response of Alto Lazio Station. The models utilize the in-house developed LEGO code, which is a modular package oriented to power plant modeling and suitable to perform transient analyses to assist during power plant design, control system design and operating procedure verification. The ability of the NSSS model to predict correctly the plant response is demonstrated through comparison with results calculated by the vendor, using REDY code, and by an in-house RETRAN-02 model

  5. Calculation of mass flow and steam quality distribution on fuel elements of light-water cooled boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Hermanns, H.J.

    1977-04-01

    By the example of light-water cooled nuclear reactors, the state of the calculation methods at disposal for calculating mass flow and steam quality distribution (sub-channel analysis) is indicated. Particular regard was paid to the transport phenomena occurring in reactor fuel elements in the range of two phase flow. Experimentally determined values were compared with recalculations of these experiments with the sub-channel code COBRA; from the results of these comparing calculations, conclusions could be drawn on the suitability of this code for defined applications. Limits of reliability could be determined to some extent. Based on the experience gained and the study of individual physical model concepts, recognized as being important, a sub-channel model was drawn up and the corresponding numerical computer code (SIEWAS) worked out. Experiments made at GE could be reproduced with the code SIEWAS with sufficient accuracy. (orig.) [de

  6. Method of providing extended life expectancy for components of boiling water reactors

    International Nuclear Information System (INIS)

    Niedrach, L.W.

    1992-01-01

    This patent describes a containment for a boiling water nuclear reactor, a stainless steel containment, the containment having a deposit of a metal of the platinum group of metals on the surfaces thereof exposed to high temperature, high pressure water of the boiling water nuclear reactor

  7. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  8. Layout of the safety analysis report for nuclear power plants with pressurized water reactor or boiling water reactor in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Albrecht, E.

    1980-01-01

    For a licence according to paragraph 7 of the Atomic Energy Act to construct and operate a nuclear power plant, the applicant has to submit a safety analysis report, which must describe the site, the plant, all hazards in connection with the plant and the proposed safety precautions. For the structure and the content of a safety analysis report, a first guideline was published in 1959. Only a few safety analysis reports were prepared nearly strictly according to this guideline. In 1976 a second guideline was published for a standard safety analysis report. The lecture deals with the guidelines. A survey over the structure and content of the German safety analysis reports will be given. The experience gained by the new safety analysis reports will be discussed. (orig.)

  9. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2005-01-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

  10. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  11. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  12. German-Brazilian nuclear deal

    International Nuclear Information System (INIS)

    Krugmann, H.

    1981-01-01

    Examination of the arguments in favor of the nuclear deal with West Germany and the resulting program suggests that revisions of both are in order to make them more compatible with Brazil's national interests. The deficiencies of current policy appear to be too weighty and numerous to be ignored. Sooner or later the government will have to move toward adjusting its nuclear agreement with West Germany, if not for the reasons discussed here then for lack of capital. Current estimates of the nuclear package lie in the range of $25 to $30 billion, compared to an initial projection of about $5 billion. The deal has become so expensive that it would draw capital from the hydropower and alcohol programs essential for the short and medium-term energy needs of the country. Mr. Krugman feels the Brazilian government should hold off on further nuclear contracts. And it should thoroughly reassess what Brazil's nuclear energy and technology requirements are and how to meet them. There are indications that the reassessment process is already underway. As long as the German nuclear industry depends on the sale of technology to Brazil, the Brazilian government will have considerable bargaining power to enforce further changes in the deal. If this power is used wisely, the result could be cooperation between the two countries toward nuclear options that are consistent with Brazil's energy and development needs

  13. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  14. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  15. Fault tolerant digital control systems for boiling water reactors

    International Nuclear Information System (INIS)

    Chakraborty, S.; Cash, N.R.

    1986-01-01

    In a Boiling Water Reactor nuclear power plant, the power generation control function is divided into several systems, each system controlling only a part of the total plant. Presently, each system is controlled by conventional analog or digital logic circuits with little interaction for coordinated control. The advent of microprocessors has allowed the development of distributed fault-tolerant digital controls. The objective is to replace these conventional controls with fault-tolerant digital controls connected together with digital communication links to form a fully integrated nuclear power plant control system

  16. Future directions in boiling water reactor design

    International Nuclear Information System (INIS)

    Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

    1987-01-01

    The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

  17. Franco-German nuclear cooperation

    International Nuclear Information System (INIS)

    Leny, J.C.; Huettl, A.

    1996-01-01

    Nuclear energy is the number one power source in the European Union. However, the first generation units would be replaced from the year 2010 onwards. In this prospect, Siemens and Framatome have drawn together in designing and commercializing a common product initially destined for the export market which has become the EPR (European Pressurized water Reactor) project. The two companies have floated with equal participation the NPI (Nuclear Power International) sub-company to manage this project. The French and German utilities participate to the financing of the project, at present at the basic stage, and the safety authorities of both countries have carried out a joint evaluation of EPR safety. With a 1500 Mwe capacity, EPR will be equipped with advanced safety systems more performing than the existing systems. Conceivers want to maintain the economic competitiveness of EPR with respect to coal power plants. EPR will take over the oldest power plants by producing a safer and cheaper energy to provide for the needs of the developed countries and then of the developing countries with no risk for the environment. An enormous effort of communication must be carried out to reduce the public anxiety and to calm down the nuclear debate and show up its merits, in particular in Europe, where its contribution is vital. (J.S.)

  18. Operating Experience in Nuclear Power Plants with Boiling-Water Reactors; Experience acquise dans l'exploitation des reacteurs a eau bouillante; Opyt ehkspluatatsii kipyashchago reaktora; Experiencia adquirida con la explotacion de reactores de agua hirviente

    Energy Technology Data Exchange (ETDEWEB)

    Ascherl, R. J. [General Electric Company, San Jose, CA (United States)

    1963-10-15

    Asignificant amount of operating experience has now been accumulated by boiling-water-reactor power plants. By the end of 1962, over 2200 million kWh of electricity have been generated by three plants operating on utility.system - Dresden Nuclear Power Station, Commonwealth Edison Company, Morris, Illinois; Vallecitos Atomic Power Plant, Pacific Gas and Electric Company and General Electric Company, Pleasanton, California; and Kahl Nuclear Power Station, Rheinisch Westfaelisches Elektrizitaetswerk and Bayernwerk, Kahl-am-Main, West Germany. Boiling-water-reactor power-plant performance, under routine electric-utility operating conditions, has been excellent. Reactor and plant availability and capacity factor provide a sound basis for anticipation of continuing reliable performance from boiling-water-reactor power stations. During 1963, four additional boiling-water-reactor plants will begin power operation: Big Rock Point Nuclear Plant, Consumers Power Company, Charlevoix, Michigan; Humboldt Bay Plant Nuclear Unit, Pacific Gas and Electric Company, Eureka, California; Garigliano Nuclear Power Station, Societa Elettronucleare Nazionale, Scauri, Italy; and Japan Power Demonstration Reactor, Japan Atomic Energy Research Institute, Tokai Mura, Japan. The start-up and initial operation of these plants confirms the expectation of reliable performance established by Dresden, Kahl, and Vallecitos. Performance records of Dresden, Kahl and Vallecitos have clearly proved the stability and safety of boiling-water reactors. Additionally, radiation levels within the plant and in the environs have been significantly below limits established by operating licences. Simplicity and ease of operation of boiling-water reactors has been confirmed. Load following characteristics of the Dresden dual-cycle boiling-water reactor have been excellent. Major and minor maintenance and repair work can be accomplished by ordinary craft unions, and without undue hardship or time limits caused by

  19. Construction of a model of the process of accumulation of radionuclides of corrosion products on the equipment in nuclear power plants with boiling-water reactors

    International Nuclear Information System (INIS)

    Tevlin, S.A.

    1985-01-01

    This paper addresses the problem of corrosion of the structural materials of the reactor loop. This problem can be solved by constructing physical models of the process of accumulation of radionuclides on the equipment at nuclear power plants and by constructing the analytical apparatus for describing them. These models are presented here, and allow the analyzing of the effect of separate states and thermophysical factors, determination of the basic factors, and the ability to foresee in timely fashion the water state and structural measures required to lower the rate of growth and to decrease the amount of radionuclides deposited on the equipment in the nuclear power plant

  20. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  1. Digital control application for the advanced boiling water reactor

    International Nuclear Information System (INIS)

    Fennern, L.E.; Pearson, T.; Wills, H.D.; Swire Rhodes, L.; Pearson, R.L.

    1986-01-01

    The Advanced Boiling Water Reactor (ABWR) is a 1300 MWe class Nuclear Power Plant whose design studies and demonstration tests are being performed by the three manufacturers, General Electric, Toshiba and Hitachi, under requirement specifications from the Tokyo Electric Power Company. The goals are to apply new technology to the BWR in order to achieve enhanced operational efficiencies, improved safety measures and cost reductions. In the plant instrumentation and control areas, traditional analog control equipment and wire cables will be replaced by distributed digital microprocessor based control units communicating with each other and the control room over fiber optic multiplexed data buses

  2. An interface redesign for the feed-water system of the advanced boiling water reactor in a nuclear power plant in Taiwan

    International Nuclear Information System (INIS)

    Hsieh Minchih; Chiu Mingchuan; Hwang Sheueling

    2014-01-01

    A well-designed human-computer interface for the visual display unit in the control room of a complex environment can enhance operator efficiency and, thus, environmental safety. In fact, a cognitive gap often exists between an interface designer and an interface user. Therefore, the issue of the cognitive gap of interface design needs more improvement and investigation. This is an empirical study that presents the application of an ecological interface design (EID) using three cases and demonstrates that an EID framework can support operators in various complex situations. Specifically, it analyzes different levels of automation and emergency condition response at the Lungmen Nuclear Power Plant in Taiwan. A simulated feed-water system was developed involving two interface styles. This study uses the NASA Task Load Index to objectively evaluate the mental workload of the human operators and the Situation Awareness Rating Technique to subjectively assess operator understanding and response, and is a pilot study investigating EID display format use at nuclear power plants in Taiwan. Results suggest the EID-based interface has a remarkable advantage over the original interface in supporting operator performance in the areas of response time and accuracy rate under both normal and emergency situations and provide supporting evidence that an EID-based interface can effectively enhance monitoring tasks in a complex environment. (author)

  3. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  4. Study of deposited crud composition on fuel surfaces in the environment of hydrogen water chemistry (HWC) of a Boiling Water Reactor at Chinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tsai, Tsuey-Lin; Lin, Tzung-Yi; Su, Te-Yen; Wen, Tung-Jen; Men, Lee-Chung

    2012-09-01

    This paper aimed at the characterization of metallic composition and surface analysis on the crud of fuel rods for unit-1 of BWR-4 at Nuclear Power Plant. The inductively coupled plasma- atomic emission spectroscopy (ICPAES) and the gamma spectrometry were carried out to analyze the corrosion product distributions and to determine the elemental compositions along the fuel rod under conditions of hydrogen water chemistry (HWC) switched from normal water chemistry (NWC) of reactor coolant in this study. Most of the crud consisted of the flakes and irregular shapes via SEM morphology. The loosely adherent oxide layer was mostly composed of hematite (α- Fe 2 O 3 ) with amorphous iron oxides by XRD results. The average deposited amounts of crud was the order of 0.5 mg/cm 2 , indicating that the fuel surface of this plant under HWC environment appeared to be one with the lower crud deposition in terms of low iron level of feedwater. It also showed no significant difference in comparison with NWC condition. (authors)

  5. 14th German nuclear law symposium

    International Nuclear Information System (INIS)

    Burgi, Martin

    2013-01-01

    Nuclear law is still relevant and topical. The nuclear power phase-out in response to the nuclear accident of Fukushima and the turnaround in German energy policy raise new legal issues. In several lectures of practioners and scientists the 14th German Nuclear Law Symposium examined questions regarding the retrofitting of nuclear power plants, their decommissioning and disposal, the current developements in the European nuclear and radiation protection law and the search for a final nuclear waste repository. The nuclear law provides examples for central challenges of administrative law, such as the independence of authorities and the protection of third parties. The discussions between the almost 150 participants are documented in several reports.

  6. Boiling water reactor stability analysis in the time domain

    International Nuclear Information System (INIS)

    Borkowski, J.A.

    1991-01-01

    Boiling water nuclear reactors may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate due to the tight coupling of flow to power, especially under gravity-driven circulation. In order to predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model has been developed for a typical boiling water reactor. Using this tool it has been demonstrated that density waxes may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases have been analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. From that study it has been concluded that two-phase friction controls the extent of the oscillation and that the existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from case to case. It has also been determined that higher dimensional nuclear feedback models reduce the extent of the oscillation. It has also been confirmed from a nonlinear dynamic standpoint that the birth of this oscillation may be described as a Hopf Bifurcation

  7. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  8. Organization of the German nuclear industry

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    Corporate ownership within the German nuclear industry has evolved constantly during the last decade, and recent acquisitions and mergers, reunification of the country, as well as preparation for a unified European power market, have led to many significant changes during the past two years. The country's nuclear industry continues to struggle under an increasingly anti-nuclear political environment, yet nuclear power provided more than one-third of Germany's total electricity generation in 1991. As in many countries, particularly in western Europe, many German companies involved in different facets of the nuclear industry are interrelated. Usually as a means of horizontal or vertical integration, the country's nuclear utilities own, directly or indirectly, shares in uranium mining projects; conversion, enrichment, and fabrication companies; or other utilities' nuclear power plants. The utilities own partial interests in companies in supporting industries as well, including transportation firms, waste management companies, uranium broker/traders, and nuclear equipment manufacturers. While the majority of the companies owned are German, numerous investments are made in non-German firms also

  9. Nonlinear dynamics of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1983-01-01

    Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

  10. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    Yoshio Matsuo

    1987-01-01

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

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

    International Nuclear Information System (INIS)

    Bottimore, R.R.

    1980-12-01

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

  12. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    Stancavage, P.

    1991-01-01

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  13. Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Lofaro, R.J.; Aggarwal, S.

    1992-01-01

    The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings

  14. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

  15. The physico-chemical 131I species in the stack exhaust air of a boiling water reactor

    International Nuclear Information System (INIS)

    Deuber, H.

    1982-07-01

    In the stack exhaust air of a German boiling water reactor, the fractions of elemental, particulate and organic 131 I were determined over a period of three years. The average fraction of elemental 131 I, which is decisive for the ingestion dose, was about 20% during the first two years and about 50% during the third year. (orig.) [de

  16. Hydrogen water chemistry for boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

    1985-01-01

    Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

  17. Nuclear energy and German foreign policy

    International Nuclear Information System (INIS)

    Kaiser, K.

    1979-01-01

    The author shows that the present foreign-policy situation with regard to the nuclear-energy complex is due to political and energy-policy origins. He is further of the opinion that the expansion of nuclear energy - in spite of internal political difficulties - will continue in the long term. Hence breeders and reconditioning will become realities. For German foreign policy this means that it must also carry responsibilities on a global scale in this respect. (orig.) [de

  18. Fifty years of German nuclear legislation

    International Nuclear Information System (INIS)

    Heller, W.

    2006-01-01

    The political situation and the state of legislation after World War II make it difficult to pinpoint a precise date of origin of German nuclear legislation. The restrictions imposed by the Allied High Commission (AHC) without any exception put a ban on the production of uranium and thorium metal as well as the construction of nuclear reactors. These restrictions were lifted expressly when the German Atomic Energy Act (AtG) entered into force on January 1, 1960, i.e. much later than the formal step of gaining sovereignty, which was marked by the protocol of May 5, 1955 terminating the Occupation Statute. In October 1955, the German federal government established the then Federal Ministry for Atomic Matters also in an attempt to reconnect to developments in nuclear fission in other parts of the Western world. To supersede the AHC law, the German federal government in December 1956 publicized a draft Atomic Energy Act. It is safe, therefore, to consider that year the starting point of German atomic legislation. This step was followed by deliberations preparatory to the adoption of the Atomic Energy Act. In 1957, however, adoption failed because no two-thirds majority was reached to amend the Basic Law, i.e. the Constitution. As a consequence, some federal states saw the need to adopt state legislation to regulate this area. On December 3, 1959, a new draft Atomic Energy Act was adopted by the German federal parliament after the second and third readings - coupled with an amendment to the Basic Law on that same day - with the votes of the opposition and with one abstention. (orig.)

  19. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  20. The German Debate on Tactical Nuclear Weapons

    International Nuclear Information System (INIS)

    Meier, Olivier

    2008-01-01

    This report analyses the debate in Germany about tactical nuclear weapons deployments in Europe. It is mainly based on interviews conducted with senior officials from the German Federal Foreign Ministry, the Federal Ministry of Defence, senior members of Parliament as well as experts from research institutes and think-tanks. The interviews focused on the more recent past in the German debate as well as the future of tactical nuclear weapon deployments in Germany and Europe. The report concludes that while a change of Germany's position on tactical nuclear weapons is unlikely to change in the short-term, several developments will make it unlikely that the continued involvement of Germany in NATO nuclear sharing will have to be debated in the medium term. Should the next Parliamentary elections, which will take place in 2009 at the latest, result in a Social Democrat-led government, a push for a reduction of Germany's involvement in NATO nuclear sharing appears possible. A conservative-led government is likely to maintain the nuclear status quo within NATO

  1. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  2. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  3. Radiation exposure in German nuclear power plants

    International Nuclear Information System (INIS)

    Mueller, W.

    1981-01-01

    The individual and collective doses in German nuclear power stations have decreased remarkably since the beginning of the commercial nuclear power production. The paper discusses the influencing factors, that have caused this development and points out areas where improvements are possible in the future. Moreover the interaction between radiation protection practice and the relevant legal regulations is considered. Usually the recording of job related doses is regarded as the most direct access to possible improvements. Concluding, it is therefore demonstrated by some examples how the evaluation of such information has taken effect in practice. (orig.) [de

  4. The availability of German nuclear power plants and possible further improvements

    International Nuclear Information System (INIS)

    Hennies, H.H.

    1985-01-01

    West Germany's 19 nuclear power plants with a capacity of 17,000 MW are actually producing about 30% of the public electricity generation. In 1984 they produced 93 billion kWh; three plants produced more than 10 billion kWh each. Most of the plants are Pressurized Water Reactors (PWR), some are Boiling Water reactors (BWR), of which the largest units went into operation recently. Considering the availability of German nuclear power plants meanwhile achieved no major systems modifications or changes in materials used are required for the time being. Instead, plant standardization is the target to be pursued. This paper discusses that the standardization of the plants allows to set up a spare part pool not just for one unit but for a parc of units. With experience of operation, maintenance and repair accumulating it is possible to have tool and repair procedures prepared well in advance for all foreseeable incidents. More optimized and automated equipment for refueling is under development to reduce even further the present annual refueling time of 30 days

  5. Invited talk on ageing management of boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    Prasad, Y.S.R.; Srinivasan, V.S.

    1994-01-01

    A nuclear power plant is built with a certain design life but by managing the operation of the plant with a well designed in-service inspection, repair and replacement programme of the equipment as required we will be able to extend the operation of the plant well beyond it's design life. This is also economically a paying proposition in view of the astronomical cost of construction of a new plant of equivalent capacity. In view of this, there is a growing trend the world over to study the ageing phenomena, especially in respect of nuclear power plant equipment and system which will contribute towards the continued operation of the nuclear power plants beyond their economic life which is fixed mainly to amortize the investments over a period. Tarapur Atomic Power Station (TAPS) which consists of 2 nos. of Boiling Water Reactor (BWRs) with the presently rated capacity of 160 MWe each has been operating for the past 24 years and is completing its 25th year of service by the year 1994 which was considered as its economic life and the plant depreciation as well as fuel supply agreement were based on this period of 25 years. I will be discussing about the available residual life which is much more than the above (25 years) and the studies we have undertaken in respect of the assessment of this residual life. (author). 2 tabs., 6 figs

  6. Great nuclear debate: German--American disagreements

    International Nuclear Information System (INIS)

    Kaiser, K.

    1978-01-01

    The post-war international consensus on nuclear nonproliferation has been eroding because of the debate over actions taken by some countries and because of the U.S. response with a unilateral change in nuclear policy. The chronology of developments is traced, noting that the 1973--74 oil crisis, India's nuclear explosion in 1974, and the exchange of Brazilian uranium for German technology all had a major role in the policy controversy. New nonproliferation proposals by the Carter administration, by precipitating debate between those wanting tighter export controls and those with nuclear commercial interests, also introduced several foreign-relations problems because of the emphasis on international agreements and a technological approach. The U.S. is credited with taking a constructive step to correct the inadequacies of the present policies and exerting pressure for global involvement in reassessing the rules on armaments and proliferation. Significant U.S. leadership could also take the form of lower energy consumption

  7. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  8. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  9. Life extension for German nuclear power plants

    International Nuclear Information System (INIS)

    Heller, W.

    2005-01-01

    The Federation of German Industries (BDI) commissioned a study of the ''Economic Effects of Alternative Lifetimes of Nuclear Power Plants in Germany.'' The expert organizations invited as authors were the Power Economy Institute of the University of Cologne (EWI) and Energy Environment Forecast Analysis GmbH (EEFA), Berlin. The reasons for commissioning the Study include the changed framework conditions (deregulation, CO 2 emission certificate trading, worldwide competition for resources), which have altered the energy supply situation in Europe. The findings of the Study were presented to the public by the BDI on October 26, 2005. The study deals with two scenarios of extended lifetimes for German nuclear power plants of 40 and 60 years as against the existing regulations with plant lifetimes limited to approx. 32 years. The longer service lives of plants are reflected in reduced electricity generation costs and thus may have a positive influence on electricity prices. Moreover, there would be additional growth of production together with additional jobs, all of which would add up to nearly 42,000 persons for all sectors of the economy as compared to the basic scenario. Also, CO 2 emissions could be curbed by up to 50 million tons of carbon dioxide. The Study offers ample and valid reasons in favor of extending the lifetimes of nuclear power plants. In the interest of general welfare, politics would be well advised to relax the restrictions on plant life in the course of this legislative term. (orig.)

  10. Availability Improvement of German Nuclear Power Plants

    International Nuclear Information System (INIS)

    Wilhelm, Oliver

    2008-01-01

    High availability is important for the safety and economical performance of Nuclear Power Plants (NPP). The strategy for availability improvement in a typical German PWR shall be discussed here. Key parameters for strategy development are plant design, availability of safety systems, component reliability, preventive maintenance and outage organization. Plant design, availability of safety systems and component reliability are to a greater extent given parameters that can hardly be influenced after the construction of the plant. But they set the frame for maintenance and outage organisation which have shown to have a large influence on the availability of the plant. (author)

  11. Ageing management in German nuclear power plants

    International Nuclear Information System (INIS)

    Becker, D.E.; Reiner, M.

    1998-01-01

    In Germany, the term 'ageing management' comprises several aspects. A demand for a special ageing monitoring programme is not explicitly contained in the regulations. However, from the Atomic Energy Act and its regulations results the operator's obligation to perform extensive measures to maintain the quality of the plant and the operating personnel working in the plant. From this point of view, comprehensive ageing management in German nuclear power plants has taken place right from the start under the generic term of quality assurance. (author)

  12. TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

    International Nuclear Information System (INIS)

    Martin, R.P.

    1993-05-01

    Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

  13. The German competence network on nuclear technology

    International Nuclear Information System (INIS)

    Kuczera, B.; Fritz, P.

    2004-01-01

    Full text: The present German energy policy is based on the phase-out of nuclear electricity generation, which means that the last of the currently operating eighteen German nuclear power plants will run until about 2022. While the plants will be shut down one after the other, decommissioning will start together with interim storage of the radioactive waste. The safe waste disposal in a final repository is planned to start around 2030 and may take another two decades, i.e., in Germany nuclear competence is further needed, at least until the mid of this century. Against this background, a high-ranking commission under the direction of the Federal Ministry of Economy and Technology evaluated the publicly funded nuclear safety related research and development (R and D) activities in Germany. One of the recommendations made by the commission was the foundation of a Competence Network on Nuclear Technology for an optimum coordination of the remaining nuclear activities including aspects of future human resources in this area. This Network was established in March 2000 with the following member institutions: Research Centre Juelich, Research Centre Karlsruhe, Research Centre Rossendorf and the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) in Munich and their neighbouring Technical Universities. The strategic objectives of the Competence Network include: Trend investigations on job development and on university education capacities in the nuclear technology sector; Enhanced cooperation of the Research Centres with universities in the nuclear field and support of international education initiatives (e.g. ENEN, WNU); Coordination and bundling of the activities in publicly funded reactor safety and waste management R and D programmes; Support of qualified young scientists and engineers (pre-doctoral students) - also by third-party funds; Participation in and collaboration with international projects and activities for advancements of international nuclear safety

  14. Recent computer applications in boiling water reactor power plants

    International Nuclear Information System (INIS)

    Hiraga, Shoji; Joge, Toshio; Kiyokawa, Kazuhiro; Kato, Kanji; Nigawara, Seiitsu

    1976-01-01

    Process computers in boiling water reactor power plants have won the position of important equipments for the calculation of the core and plant performances and for data logging. Their application technique is growing larger and larger every year. Here, two systems are introduced; plant diagnostic system and computerized control panel. The plant diagnostic system consists of the part processing the signals from a plant, the operation part mainly composed of a computer to diagnose the operating conditions of each system component using input signal, and the result display (CRT or typewriter). The concept on the indications on control panels in nuclear power plants is changing from ''Plant parameters and to be indicated on panel meters as much as possible'' to ''Only the data required for operation are to be indicated.'' Thus the computerized control panel is attracting attention, in which the process computer for processing the operating information and CRT display are introduced. The experimental model of that panel comprises and operator's console and a chief watchmen's console. Its functions are dialogic data access and the automatic selection of preferential information. (Wakatsuki, Y.)

  15. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

  16. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  17. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    International Nuclear Information System (INIS)

    Gavilian-Moreno, Carlos; Espinosa-Paredes, Gilberto

    2016-01-01

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution

  18. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gavilian-Moreno, Carlos [Iberdrola Generacion, S.A., Cofrentes Nuclear Power Plant, Project Engineering Department, Paraje le Plano S/N, Valencia (Spain); Espinosa-Paredes, Gilberto [Area de ingeniera en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Mexico city (Mexico)

    2016-04-15

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  19. Using Largest Lyapunov Exponent to Confirm the Intrinsic Stability of Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Carlos J. Gavilán-Moreno

    2016-04-01

    Full Text Available The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs. Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  20. Safety analysis for boiling water reactors

    International Nuclear Information System (INIS)

    Kersting, E.; Linden, J. von; Mueller-Ecker, D.; Werner, W.

    1993-07-01

    This report is the translation of GRS-95 'Sicherheitsanalyse fuer Siedewasserreaktoren - Zusammenfassende Darstellung'. Recent analysis results -concerning the chapters on accident management, fire and earthquake - that were not included in the German text have been added to this translation. In cases of doubt, GRS-102 (main volume) is the factually correct version. (orig.)

  1. Passive gamma analysis of the boiling-water-reactor assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  2. German nuclear fuel exports and imports 1991

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The statistics compiled by the German Federal Office for Trade and Industry (Bundesamt fuer Wirtschaft) for the Federal Ministry for the Environment, Conservation of Nature, and Reactor Safety of imports and exports of nuclear fuels and source materials in 1991 show a major drop by 33.8% in imports and a pronounced rise by 191.5% in exports, compared to the levels in the previous year. Source material for the purposes of these statistics refers only to uranium concentrate. Quantitatively, the biggest import items are source materials, depleted uranium, and uranium enriched up to 3%. Exports of unirradiated material quantitatively comprise mainly depleted uranium, source material, and uranium enriched up to 10%. (orig.) [de

  3. The effects of aging on Boiling Water Reactor core isolation cooling system

    International Nuclear Information System (INIS)

    Lee, Bom Soon.

    1994-01-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes

  4. Study and application of boiling water reactor jet pump characteristic

    International Nuclear Information System (INIS)

    Liao Lihyih

    1992-01-01

    RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model. In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible. In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power, level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K d , the suction flow junction form loss coefficients, K s , the diffuser form loss coefficient, K c , and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve. (orig.)

  5. Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

    2008-08-15

    AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

  6. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    Queiser, H.

    1976-01-01

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.) [de

  7. Assessment of accident risks from german nuclear plants

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1979-01-01

    The German risk study are presented. The main objectives can be summed up as follows: (a) An assessment of the societal risk due to accidents in nuclear power plants with reference to German conditions; (b) To get experience in the field of risk analysis and to provide a basis for estimation of uncertainties; (c) To provide guidance for future activities in the German Reactor Safety Research Program. Finally several conclusions reached by this study are discussed. (author)

  8. The German-German history of the nuclear power plant Greifswald. Nuclear power between east and west. 2. ed.

    International Nuclear Information System (INIS)

    Hoegselius, Per

    2015-01-01

    The historical study covers the chapters The nuclear power plant Greifswald; Lubmin shortly before the ''Wende'' 1989; the German ''Wende''; from the last vote for the ''Volkskammer (parliament of the German Democratic Republic) to the German reunification; Lubmin in reunified Germany; conclusions and perspectives. In the attachment technical data about the reactors WWER-440/W-230 are summarized, including a list of WWERs in the former eastern bloc countries.

  9. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2001-06-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current nuclear applications Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for the purpose. The present volume describes the specification of such a benchmark. The transient addressed is a turbine trip (TT) in a BWR involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the plant make the present benchmark very valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

  10. Electrochemical measurements and modeling predictions in boiling water reactors under various operating conditions

    International Nuclear Information System (INIS)

    Indig, M.E.

    1991-01-01

    One important issue for providing life extension to operating boiling water nuclear reactors (BWRs) is the control of stress corrosion cracking in all sections of the primary coolant circuit. This paper links experimental and theoretical methods that provide understanding and measurements of the critical parameter, the electrochemical potential (ECP), and its application to determining crack growth rate among and within the family of BWRs. Measurement of in-core ECP required the development of a new family of radiation-resistant sensors. With these sensors, ECPs were measured in the core and piping of two operating BWRs. Concurrent crack growth measurements were used to benchmark a crack growth prediction algorithm with measured ECPs

  11. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  12. Calculation of optimum control rod operation programme for boiling water reactor

    International Nuclear Information System (INIS)

    Fehr, L.

    1978-01-01

    Control rod operation programmes are calculated based on a three dimensional Boiling Water Reactor situation model. The position of the control rods at variosu burn-ups is chosen by an optimisation so that the sum of the square deviations of the load density distribution from an optimum distribution ('Haling' distribution) are minimised. Other conditions are remaining critical and observing the thermal limits for central fuel element melting and critical heat surface loading. As an example, an optimum control rod operation programme for the first cycle in Lengen nuclear power station is calculated and is compared with the programme actually used. (orig.) 891 HP [de

  13. Power distribution effects on boiling water reactor stability

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.

    1989-01-01

    The work presented in this paper deals with the effects of spatial power distributions on the stability of boiling water reactors (BWRs). It is shown that a conservative power distribution exists for which the stability is minimal. These results are relevant because they imply that bounding stability calculations are possible and, thus, a worst-possible scenario may be defined for a particular BWR geometry. These bounding calculations may, then, be used to determine the maximum expected limit-cycle peak powers

  14. Calculation of limit cycle amplitudes in commercial boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Perez, R.B.; Cacuci, D.G.

    1984-01-01

    This paper describes an investigation of the dynamic behavior of a boiling water reactor (BWR) in the nonlinear region corresponding to linearly unstable conditions. A nonlinear model of a typical BWR was developed. The equations underlying this model represent a one-dimensional void reactivity feedback, point kinetics with a single delayed neutron group, fuel behavior, and recirculation loop dynamics (described by a single-node integral momentum equation)

  15. Models and Stability Analysis of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Dorning, John

    2002-01-01

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  16. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  17. Development of advanced boiling water reactor for medium capacity

    International Nuclear Information System (INIS)

    Kazuo Hisajima; Yutaka Asanuma

    2005-01-01

    This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

  18. German nuclear law day 2004 - a conference report

    International Nuclear Information System (INIS)

    Anon.

    2005-01-01

    Topical issues of nuclear law constituted the main subjects discussed at the 2004 German Nuclear Law Day organized in Berlin on November 11 to 12, 2004. The agenda included actual issues potentially arising from the topics final storage of nuclear waste, financing a new site search for a repository, and supervision of nuclear installations. Experts from the administration of justice, the federal and state governments, law offices, universities, and the industry discussed the matters in 14 lectures. (orig.)

  19. Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

    International Nuclear Information System (INIS)

    Wheeler, J.K.; Pallotta, A.S.

    1992-01-01

    The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

  20. Steam generation: fossil-fired systems: utility boilers; industrial boilers; boiler auxillaries; nuclear systems: boiling water; pressurized water; in-core fuel management; steam-cycle systems: condensate/feedwater; circulating water; water treatment

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    A survey of development in steam generation is presented. First, fossil-fired systems are described. Progress in the design of utility and industrial boilers as well as in boiler auxiliaries is traced. Improvements in coal pulverizers, burners that cut pollution and improve efficiency, fans, air heaters and economisers are noted. Nuclear systems are then described, including the BWR and PWR reactors, in-core fuel management techniques are described. Finally, steam-cycle systems for fossil-fired and nuclear power plants are reviewed. Condensate/feedwater systems, circulating water systems, cooling towers, and water treatment systems are discussed

  1. The physico-chemical 131I species in the exhaust air of a boiling water reactor (BWR 4)

    International Nuclear Information System (INIS)

    Deuber, H.

    1982-12-01

    In a German boiling water reactor, the physico-chemical 131 I species were determined in the plant exhaust and in the individual exhausts during 12 months. These measurements aimed in particular at determining the percentage and the source of the radiologically decisive elemental 131 I released to the environment. The retention of the 131 I species by iodine filters was also investigated. On an average, 45% of the 131 I discharged with the plant exhaust consisted of elemental iodine. This was largely released with the exhaust from the reactor building and from the turbine building. The other 55% consisted almost entirely of organic I. (orig./HP) [de

  2. The physico-chemical I-131 species in the exhaust air of a boiling water reactor (BWR 5)

    International Nuclear Information System (INIS)

    Deuber, H.

    1984-02-01

    In a German boiling water reactor, the pysico-chemical I-131 species were determined in the plant exhaust and in the individual exhausts during four months. These measurements aimed in particular at determining the percentage and the source of the radiologically decisive elemental I-131 released to the environment. On an average 13% of the I-131 discharged with the plant exhaust consisted of elemental iodine. This was largely released with the exhausts from the reactor building and from the turbine building. The main component was organic-bound I. (orig./HP) [de

  3. A review of boiling water reactor water chemistry: Science, technology, and performance

    International Nuclear Information System (INIS)

    Fox, M.J.

    1989-02-01

    Boiling water reactor (BWR) water chemistry (science, technology, and performance) has been reviewed with an emphasis on the relationships between BWR water quality and corrosion fuel performance, and radiation buildup. A comparison of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.56, the Boiling Water Reactor Owners Group (BWROG) Water Chemistry Guidelines, and Plant Technical Specifications showed that the BWROG Guidelines are more stringent than the NRC Regulatory Guide, which is almost identical to Plant Technical Specifications. Plant performance with respect to BWR water chemistry has shown dramatic improvements in recent years. Up until 1979 BWRs experienced an average of 3.0 water chemistry incidents per reactor-year. Since 1979 the water chemistry technical specifications have been violated an average of only 0.2 times per reactor-year, with the most recent data from 1986-1987 showing only 0.05 violations per reactor-year. The data clearly demonstrate the industry-wide commitment to improving water quality in BWRs. In addition to improving water quality, domestic BWRs are beginning to switch to hydrogen water chemistry (HWC), a remedy for intergranular stress corrosion cracking. Three domestic BWRs are presently operating on HWC, and fourteen more have either performed HWC mini tests or are in various stages of HWC implementation. This report includes a detailed review of HWC science and technology as well as areas in which further research on BWR chemistry may be needed. 43 refs., 30 figs., 8 tabs

  4. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

  5. Domestic and overseas development of advanced boiling water reactors

    International Nuclear Information System (INIS)

    Hatazawa, Mamoru; Fuchino, Satoshi; Nakada, Kotaro

    2010-01-01

    Since Toshiba delivered the world's first advanced boiling water reactor (ABWR) to The Tokyo Electric Power Company, Inc. in 1996, we have been devoting continuous efforts to the construction and operational support of ABWR systems as major products. We are now promoting the construction of domestic and overseas ABWR systems along with the standardization of ABWRs. We are also engaged in the research and development of core technologies to support further promotion of ABWRs as a concurrent solution to the issues of global warming and energy security for individual countries. (author)

  6. A model for oxidizing species concentrations in boiling water reactors

    International Nuclear Information System (INIS)

    Sun, B.; Chexal, B.; Pathania, R.; Chun, J.; Ballinger, R.; Abdollahian, D.

    1993-01-01

    To evaluate and control the intergranular stress corrosion cracking of boiling water reactor (BWR) vessel internal components requires knowledge of the concentration of oxidizing species that affects the electrochemical potentials in various regions of a BWR. In a BWR flow circuit, as water flows through the radiation field, the radiolysis process and chemical reactions lead to the production of species such as oxygen, hydrogen, and hydrogen peroxide. Since chemistry measurements are difficult inside BWRs, analytical tools have been developed by Ruiz and Lin, Ibe and Uchida and Chun and Ballinger for estimating the concentration of species that provide the necessary input for water chemistry control and material protection

  7. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  8. Control rooms in German nuclear power plants

    International Nuclear Information System (INIS)

    Hoffmann, E.

    1999-01-01

    The paper explains and illustrates the dissimilarity in design and equipment of control rooms in German NPPs, as well as a historical survey of the general principles and approaches applied in the evolution of control room technology, including backfitting activities. Experience obtained from daily operation as well training at the simulators is taken as a basis to formulate fundamental requirements for modification or novel design approaches. (orig./CB) [de

  9. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  10. Fuel performance in the Barsebeck boiling water reactors (Unit 1 and 2)

    International Nuclear Information System (INIS)

    Norman, B.

    1979-01-01

    Sydkraft is the largest privately owned utility in Sweden. It serves about 20% of the Swedish population with about 12 TWh of electric power per year, of which 64% is nuclear (1978 figures). The two identical 590 MWE ASEA-ATOM boiling water reactors in Barsebeck have been in operation since 1975 and 1977 respectively. Fission product activity in the primary circuits and in the off-gas systems is extremely low and indicate a near perfect fuel condition. Operating restrictions limiting the effect of pellet cladding interaction have been in use since initial start-up and testing. A few events involving rapid power increases above the preconditioned power level have occurred without causing fuel failures. It is believed that an analysis of power reactor operational transients, which did not cause fuel failures, can be useful to design more adequate and less conservative rules for the operation of nuclear reactor cores

  11. Aging assessment of the boiling-water reactor (BWR) standby liquid control system

    International Nuclear Information System (INIS)

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious

  12. evelopment of a boiling water reactor fault diagnostic system with a signed directed graph method

    International Nuclear Information System (INIS)

    Chen, M.; Yu, C.C.; Liou, C.T.; Liao, L.Y.

    1990-01-01

    The fault diagnostic system for a nuclear power reactor is expected to be a useful decision support system for the operators during transients and accident conditions. A considerable research effort has been devoted to the development of automated fault diagnostic systems. One major approach, which has been widely used in chemical engineering, is to identify the possible causes of process disturbance using a logic-oriented method called signed directed graph (SDG). A knowledge based system was developed with the rules derived from the SDG representation. The SDG for the Chinshan nuclear power plant, which is a typical boiling water reactor, is established. The personal consultant system is used as the expert system development tool in this paper

  13. Using NJOY99 and MCNP4B2 to Estimate the Radiation Damage Displacements per Atom per Second in Steel Within the Boiling Water Reactor Core Shroud and Vessel Wall from Reactor-Grade Mixed-Oxide/Uranium Oxide Fuel for the Nuclear Power Plant at Laguna Verde, Veracruz, Mexico

    International Nuclear Information System (INIS)

    Vickers, Lisa

    2003-01-01

    The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased 239 Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s -1 ) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s -1 ) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor

  14. Educating nuclear engineers at German universities

    International Nuclear Information System (INIS)

    Knorr, J.

    1995-01-01

    Nuclear technology is a relatively young university discipline. Yet, as a consequence of the declining public acceptance of the peaceful use of nuclear power, its very existence is already being threatened at many universities. However, if Germany needs nuclear power, which undoubtedly is the case, highly qualified, committed experts are required above all. Nuclear technology develops internationally. Consequently, also university education must meet international standards. Generally, university education has been found to be the most effective way of increasing the number of scientific and engineering personnel. Nuclear techniques have meanwhile found acceptance in many other scientific disciplines, thus advancing those branches of science. Teaching needs research; like research in nucelar technology at the national research centers, also the universities are suffering massive financial disadvantages. Research is possible only if outside funds are solicited, which increase dependency and decreases basic research. (orig.) [de

  15. Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

    International Nuclear Information System (INIS)

    Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

    1990-11-01

    The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs

  16. German nuclear energy development and international cooperation

    International Nuclear Information System (INIS)

    Schmidt-Kuster, W.J.

    1985-01-01

    The author gives a brief survey on the short, but relatively successful story of nuclear energy in the Federal Republic of Germany (FRG). Like many other countries, FRG had to go through a very difficult period of political indecision and violent opposition from antinuclear groups, supported by large parts of the media. The licensing procedures have been streamlined, nuclear power plants are being built without major interference, and the FRG is making good progress in closing the fuel cycle. This means that nuclear power will play an important role in the energy supply system, although on a lower level than originally anticipated

  17. 10th German nuclear law symposium

    International Nuclear Information System (INIS)

    Koch, H.J.; Rossnagel, A.

    2000-01-01

    This 10th symposium on nuclear law in Germany was held eight years after the 9th symposium. Due to the change of government after the last general elections, there had been a turnaround in Germany's energy policy. 'Phasing out nuclear energy' was the major strategy of the new Federal Government. The topics of the papers presented at the symposium therefore focus on: a new time frame for NPP shutdown and termination of operating licences; ensuring the safe operation of nuclear power plants for the remaining operating periods; new concepts for radwaste management and ultimate disposal. (orig./CB) [de

  18. Losses in German nuclear power plants

    International Nuclear Information System (INIS)

    Abinger, R.

    1982-01-01

    The author illustrates the special features of engineering insurance for nuclear power plants. The shares of the Allianz Versicherungs-AG in the insurance of construction and erection work and in machinery insurance are dealt with. Risk estimation is usually based on statistical analysis of losses. Loss analysis in the conventional sector of nuclear power plants shows typical characteristics of traditional erection and machinery losses. In the nuclear field, however, costs are greatly increased by added safety measures. For this reason, additional cover is allocated and incorporated in premium assessment. Examples from erection and machinery reveal the greater costs involved in handling losses. (orig.) [de

  19. Teachers and nuclear energy - German situation

    International Nuclear Information System (INIS)

    Schaeferhenrich, B.

    1994-01-01

    School education in Germany is the responsibility of each federal state; there is no national curriculum. Nevertheless, all pupils of all types of school are confronted with the energy topic; nuclear energy is a compulsory topic in a subject where pupils' achievements are marked; interdisciplinary thinking is encouraged. The approach may however fail because of inflexible teachers (against nuclear energy), organizational structures and complementary training

  20. A decision support system for maintenance management of a boiling-water reactor power plant

    International Nuclear Information System (INIS)

    Shen, J.H.; Ray, A.; Levin, S.

    1996-01-01

    This article reports the concept and development of a prototype expert system to serve as a decision support tool for maintenance of boiling-water reactor (BWR) nuclear power plants. The code of the expert system makes use of the database derived from the two BWR units operated by the Pennsylvania Power and Light Company in Berwick, Pennsylvania. The operations and maintenance information from a large number of plant equipment and sub-systems that must be available for emergency conditions and in the event of an accident is stored in the database of the expert system. The ultimate goal of this decision support tool is to identify the relevant Technical Specifications and management rules for shutting down any one of the plant sub-systems or removing a component from service to support maintenance. 6 refs., 7 figs

  1. A symptom based decision tree approach to boiling water reactor emergency operating procedures

    International Nuclear Information System (INIS)

    Knobel, R.C.

    1984-01-01

    This paper describes a Decision Tree approach to development of BWR Emergency Operating Procedures for use by operators during emergencies. This approach utilizes the symptom based Emergency Procedure Guidelines approved for implementation by the USNRC. Included in the paper is a discussion of the relative merits of the event based Emergency Operating Procedures currently in use at USBWR plants. The body of the paper is devoted to a discussion of the Decision Tree Approach to Emergency Operating Procedures soon to be implemented at two United States Boiling Water Reactor plants, why this approach solves many of the problems with procedures indentified in the post accident reviews of Three Mile Island procedures, and why only now is this approach both desirable and feasible. The paper discusses how nuclear plant simulators were involved in the development of the Emergency Operating Procedure decision trees, and in the verification and validation of these procedures. (orig./HP)

  2. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  3. Boiling water reactor containment modeling and analysis at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Holcomb, E.E. III; Wilson, G.E.

    1984-01-01

    Under the auspices of the United States Nuclear Regulatory Commission, severe accidents are being studied at the Idaho National Engineering Laboratory. The boiling water reactor (BWR) studies have focused on postulated anticipated transients without scram (ATWS) accidents which might contribute to severe core damage or containment failure. A summary of the containment studies is presented in the context of the analytical tools (codes) used, typical transient simulation results and the need for prototypical containment data. All of these are related to current and future analytical capabilities. It is shown that torus temperatures during the ATWS depart from limiting conditions for BWR T-quencher operation, outside of which stable steam condensation has not been proven

  4. Final environmental statement for La Crosse Boiling Water Reactor: (Docket No. 50-409)

    International Nuclear Information System (INIS)

    1980-04-01

    A Final Environmental Statement for the Dairyland Power Cooperative for the conversion from a provisional to a full-term operating license for the La Crosse Boiling Water Reactor, located in Vernon County, Wisconsin, has been prepared by the Office of Nuclear Reactor Regulation. This statement provides a summary of environmental impacts and adverse effects of operation of the facility, and a consideration of principal alternatives (including removal of LACBWR from service, alternative cooling methodology, and alternative waste treatment systems). Also included are the comments of federal, state, and local governmental agencies and certain non-governmental organizations on the La Crosse Draft Environmental Statement and staff responses to these comments. After weighing environmental, economic, and technical benefits and liabilities, the staff recommends conversion from a provisional operating license to a full-term operating license, subject to specific environmental protection limitations. An operational monitoring program shall be established as part of the Environmental Technical Specifications. 64 refs., 20 figs., 48 tabs

  5. Analysis of boiling water reactors capacities for the 100% MOX fuel recycling

    International Nuclear Information System (INIS)

    Knoche, Dietrich

    1999-01-01

    The electro-nuclear park exploitation leads to plutonium production. The plutonium recycling in boiling water reactors performs a use possibility. The difference between the neutronic characteristics of the uranium and the plutonium need to evaluate the substitution impact of UOX fuel by MOX fuel on the reactor operating and safety. The analysis of the main points reached to the following conclusions: the reactivity coefficients are negative, during a cooling accident the re-divergence depends on the isotopic vector of the used plutonium, the efficiency lost of control cross resulting from the plutonium utilization can be compensate by the increase of the B 4C enrichment by 10 B and the change of the steel structure by an hafnium structure, the reactivity control in evolution can be obtained by the fuel poisoning (gadolinium, erbium) and the power map control by the plutonium content monitoring. (A.L.B.)

  6. Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

    2004-11-01

    One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

  7. Analysis of the dynamics of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    Castillo D, R.

    1996-01-01

    The March-Leuba lineal reduced model is represented mathematically by a differential equations system, which corresponds to the direct transfer function, punctual kinetics approximation, neutron field dynamics, heat transfer in fuels, and channel dynamics approximation that relates the fuel temperature changes to the reactivity changes by vacuums. The model presents significant differences in one of the equation coefficients. The Pade order approximation used for the equation deduction for the channel has a different behavior to the exponential one for long periods of bubble residence. (Author)

  8. The interpretation of neutron noise in boiling water reactors

    International Nuclear Information System (INIS)

    John, T.M.; Singh, O.P.

    1985-01-01

    Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

  9. Advanced boiling water reactors for the 90's and beyond

    International Nuclear Information System (INIS)

    Rao, A.S.; Sawyer, C.D.; Qurik, J.F.; McCandless, R.J.

    1990-01-01

    This paper discusses how the advanced boiling water reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability and reduced occupational exposure and radwaste. International cooperative efforts are also under way aimed at development of a simplified BWR employing natural circulation and passive safety systems. The SBWR conceptual design is complete. This BWR concept shows technical and economic promise. The SBWR program is aimed at providing a U.S. NRC certified design in an investor-ready state by 1995. With its short construction schedule, the 600 MWe SBWR will provide an option for commercial operation worldwide by the mid-to-late 1990s

  10. Spectral shift rod for the boiling water reactor

    International Nuclear Information System (INIS)

    Yokomizo, O.; Kashiwai, S.; Nishida, K.; Orii, A.; Yamashita, J.; Mochida, T.

    1993-01-01

    A Boiling Water Reactor core concept has been proposed using a new fuel component called spectral shift rod (SSR). The SSR is a new type of water rod in which a water level is formed during core operation. The water level can be controlled by the core recirculation flow rate. By using SSRs, the reactor can be operated with all control rods withdrawn through the operation cycle as well as that a much larger natural uranium saving is possible due to spectral shift operation than in current BWRs. The steady state and transient characteristics of the SSRs have been examined by experiments and analyses to certify the feasibility. In a reference design, a four times larger spectral shift width as for the current BWR has been obtained. (orig.)

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

    International Nuclear Information System (INIS)

    Fukutomi, S.; Takigawa, Y.

    1992-01-01

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

  12. Experiences in stability testing of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Otaduy, P.J.

    1986-01-01

    The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

  13. Boiling water reactor liquid radioactive waste processing system

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

  14. Expert system for control rod programming of boiling water reactors

    International Nuclear Information System (INIS)

    Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

    1986-01-01

    Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

  15. Boiling water reactor modeling capabilities of MMS-02

    International Nuclear Information System (INIS)

    May, R.S.; Abdollahian, D.A.; Elias, E.; Shak, D.P.

    1987-01-01

    During the development period for the Modular Modeling System (MMS) library modules, the Boiling Water Reactor (BWR) has been the last major component to be addressed. The BWRX module includes models of the reactor core, reactor vessel, and recirculation loop. A pre-release version was made available for utility use in September 1983. Since that time a number of changes have been incorporated in BWRX to (1) improve running time for most transient events of interest, (2) extend its capability to include certain events of interest in reactor safety analysis, and (3) incorporate a variety of improvements to the module interfaces and user input formats. The purposes of this study were to briefly review the module structure and physical models, to point the differences between the MMS-02 BWRX module and the BWRX version previously available in the TESTREV1 library, to provide guidelines for choosing among the various user options, and to present some representative results

  16. Corrosion problems in boiling water reactors and their remedies

    International Nuclear Information System (INIS)

    Rosborg, B.

    1989-01-01

    This article briefly presents current corrosion problems in boiling water reactors and their remedies. The problems are different forms of environmentally assisted cracking, and the remedies are divided into material-, environment-, and stress-related remedies. The list of problems comprises: intergranular stress corrosion cracking (IGSCC) in weld-sensitized stainless steel piping; IGSCC in cold-bent stainless steel piping; irradiation-assisted stress corrosion cracking (IASCC) in stainless alloys; IGSCC in high-strength stainless alloys. A prospective corrosion problem, as judged from literature references, and one which relates to plant life, is corrosion fatigue in pressure vessel steel, since the reactor pressure vessel is the most critical component in the BWR pressure boundary as regards plant safety. (author)

  17. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    PetrusTakaki, N.

    2012-01-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  18. Analysis of scrams and forced outages at boiling water reactors

    International Nuclear Information System (INIS)

    Earle, R.T.; Sullivan, W.P.; Miller, K.R.; Schwegman, W.J.

    1980-07-01

    This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability

  19. Age-related degradation of boiling water reactor vessel internals

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1992-01-01

    Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

  20. Outline of the advanced boiling water reactor (ABWR)

    International Nuclear Information System (INIS)

    Hucik, S.A.; Imaoka, T.; Minematsu, A.; Takashima, Y.

    1986-01-01

    The fundamental design of the Advanced Boiling Water Reactor (ABWR) was completed in December 1985. This design represents the next generation of Boiling Water Reactors (BWR) to be introduced into commercial operation in the 1990s. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technologies and many new improvements based on an extensive accumulation of world-wide experience through design, construction and operation of BWRs. The ABWR development program was initiated in 1978, with subsequent design and test and development programs started in 1981. Most of the development and verification tests of the new features have been completed. The ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. The ABWR objectives were specific improvements such as operating and safety margins, enhanced availability and capacity factor, and reduced occupational exposure while at the same time achieving significant cost reduction in both capital and operating costs. The ABWR is characterized by an improved NSSS including ten internal recirculation pumps, fine motion electric-hydraulic control rod drives, optimized safety and auxiliary systems, advanced control and instrumentation systems, improved turbine-generator with moisture/separator reheater with plant output increased to 1350 MWe, and an integrated reinforced concrete containment vessel and compact Reactor and Turbine Building design. The turbine system also included improvements in the Turbine-Generator, feedwater/heater system, and condensate treatment systems. The radwaste system was also optimized taking advantage of the plant design improvements and advances in radwaste technology. The ABWR is a truly optimal design which utilizes advanced technologies, capabilities, performance improvements, and yet provides an economic advantage. (author)

  1. German nuclear expansion: state, capital, world market

    International Nuclear Information System (INIS)

    Galvan, C.G.

    1988-01-01

    This paper intends to discuss the technological development as it happened in Germany or, better, it places in the scene of world market, where it did. In the attention center is the big achievement of pacific use of atomic technology: the nuclear power plants, which the new energy is used in electric generation. (C.M.)

  2. Radioactive waste management for German nuclear power plants

    International Nuclear Information System (INIS)

    Weh, R.; Methling, D.; Sappok, M.

    1996-01-01

    In Germany, back-end fuel cycle provisions must be made for the twenty nuclear power plants currently run by utilities with an aggregate installed power of 23.4 GWe, and the four nuclear power plants already shut down. In addition, there are the shut down nuclear power plants of the former German Democratic Republic, and a variety of decommissioned prototype nuclear power plants built with the participation of the federal government and by firms other than utilities. The nuclear power plants operated by utilities contribute roughly one third of the total electricity generation in public power plants, thus greatly ensuring a stable energy supply in Germany. The public debate in Germany, however, focuses less on the good economic performance of these plants, and the positive acceptance at their respective sites, but rather on their spent fuel and waste management which, allegedly, is not safe enough. The spent fuel and waste management of German nuclear power plants is planned on a long-term basis, and executed in a responsible way by proven technical means, in the light of the provisions of the Atomic Act. Each of the necessary steps of the back end of the fuel cycle is planned and licensed in accordance with German nuclear law provisions. The respective facilities are built, commissioned, and monitored in operation with the dedicated assistance of expert consultants and licensing authorities. Stable boundary conditions are a prerequisite in ensuring the necessary stability in planning and running waste management schemes. As producers of waste, nuclear power plants are responsible for safe waste management and remain the owners of that waste until it has been accepted by a federal repository. (orig./DG) [de

  3. German nuclear policy reconsidered. Implications for the electricity market

    International Nuclear Information System (INIS)

    Fuersch, Michaela; Lindenberger, Dietmar; Malischek, Raimund; Nagl, Stephan; Panke, Timo; Trueby, Johannes

    2011-01-01

    In the aftermath of the nuclear catastrophe in Fukushima, German nuclear policy has been reconsidered. This paper demonstrates the economic effects of an accelerated nuclear phase-out on the German electricity generation sector. A detailed optimization model for European electricity markets is used to analyze two scenarios with different lifetimes for nuclear plants (phase-out vs. prolongation). Based on political targets, both scenarios assume significant electricity demand reductions and a high share of generation from renewable energy sources in Germany. Our principal findings are: First, nuclear capacities are mainly replaced by longer lifetimes of existing coal-fired plants and the construction of new gas-fired plants. Second, fossil fuel-based generation and power imports increase, while power exports are reduced in response to the lower nuclear generation. Third, despite the increased fossil generation, challenging climate protection goals can still be achieved within the framework of the considered scenarios. Finally, system costs and electricity prices are clearly higher. We conclude that the generation sector can generally cope with an accelerated nuclear phase-out under the given assumptions. Yet, we emphasize that such a policy requires a substantial and costly transformation of the supply and the demand side.

  4. Casualties in Federal German nuclear facilities

    International Nuclear Information System (INIS)

    1989-01-01

    The social insurance companies are responsible for compensation in case of occupational accidents and therefore keep statistics on occupational accidents, which for the workplace 'nuclear installations' show the following data covering the period 1st January 1969 up to 1988: 25 fatal occupational accidents; 1 fatal occupational disease; 20 fatal road accidents. The 25 occupational accidents were of the conventional type. The death in 1988 caused by occupational disease was due to an asbestos mesothelioma acquired by the person during its former occupation in a coal-fired power plant. (orig./HP) [de

  5. The German nuclear power plant safety study

    International Nuclear Information System (INIS)

    1979-01-01

    With this study a new approach has been chosen, taking nuclear power plants as an example to assess and to describe the risks arising from the use of modern technology, including those hazards emanating from the rather hypothetical possibility of occurrence of very serious accidents. Following the definition of basic concepts and methods to be applied in risk assessment studied, as well as a brief account of the design and operating mode of nuclear power plants with PWRs', accidents and failures to be considered in a safety study are described. Using the course-of-event and fault tree analysis, the probability of fission product release as a consequence of failures in safety systems or of core meltdown is evaluated. Subsequently, the theoretical model for assessment of reactor accident consequences is presented, discussing such aspects as the dispersion of radioactivity in the atmosphere, the radiation dose model, safety and countermeasures, the model for the evaluation of health hazards as well as methods and calculations for estimating the reliability of risk assessments together with the remaining uncertainties. In an appendix to this study, the analyses presented in the study are discussed in the light of the TMI-2 event. This safety study showing the possibilities of detecting, keeping in check and minimizing harmful effects, can be regarded as a contribution to a better understanding of our modern, highly industrialised society, and eventually to an improvement of the quality of life. (GL) 891 GL/GL 892 MB [de

  6. The German risk study for nuclear power plants

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1980-01-01

    In August 1979 results of the ''German Risk Study for Nuclear Power Plants'' were published. The Main Report, in which approach and results of the study are documented, has been available since the end of 1979. It was the charter of the study - which was performed on behalf of the Minister of Research and Technology of the Federal Republic of Germany - to apply as far as possible the methods of the US Reactor Safety Study (WASH-1400) to German plant and site conditions. A direct transfer of the results was not deemed justified, mainly for the following reasons: There is quite a number of differences between the design of the reference plants of WASH-1400 (Surry-1, Peach Bottom-2) and German nuclear power plants. The mean population density in the Federal Republic of Germany is more than ten times of the United States. In the vicinity of nuclear power plants the ratio is about 3:1. To calculate the collective risk resulting from reactor accidents, a total of 25 plants at 19 different sites in the Federal Republic of Germany were considered. This included all plants with 600 MW or more electrical output, which were in operation, under construction or in licensing process by July 1, 1977. As an approximation to the real situation, it has been assumed that all 25 plants are technically identical to the reference plant

  7. Pressurized thermal shock analysis in German nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Fricke, Stefan; Braun, Michael [TUEV NORD Nuclear, Hannover (Germany)

    2015-03-15

    For more than 30 years TUeV NORD is a competent consultant in nuclear safety is-sues giving expert third party opinion to our clients. According to the German regulations the safety against brittle fracture has to be proved for the reactor pressure vessel (RPV) and with a new level of knowledge the proof has to be continuously updated with the development in international codes and standards like ASME, BS and RCC-M. The load of the RPV is a very complex transient pressure and temperature situation. Today these loading conditions can be modeled by thermal hydraulic calculations and new experimental results much more detailed than in the construction phase of German Nuclear Power Plants in the 1980s. Therefore, the proof against brittle fracture from the construction phase had to be updated for all German Nuclear Power Plants with the new findings of the loading conditions especially for a postulated small leakage in the main coolant line. The RPV consists of ferritic base material (about 250 mm) and austenitic cladding (about 6 mm) at the inner side. The base material and the cladding have different physical properties which have to be considered temperature dependently in the cal-culations. Radiation-embrittlement effects on the material are to be respected in the fracture mechanics assessment. The regions of the RPV of special interest are the core weld, the inlet and outlet nozzle region and the flange connecting weld zone. The fracture mechanics assessment is performed for normal and abnormal operating conditions and for accidents like LOCA (Loss of Coolant Accident). In this paper the German approach to fracture mechanics assessment to brittle fracture will be discussed from the point of view of a third party organization.

  8. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Boing, L.E.; Henley, D.R.; Manion, W.J.; Gordon, J.W.

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs

  9. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  10. German risk study 'nuclear power plants, phase B'

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1989-01-01

    The results of the German risk study 'Nuclear power plants, phase B' indicate that an accident in a nuclear power plant which cannot be managed by the safety systems according to design, is extremely improbable: Its probability is at about 3 to 100,000 per year and plant. Even if the safety systems fail, emergency measures can be effected in a nuclear power plant to prevent an accident. These in-plant emergency measures diminish the probability of a core meltdown to about 4 to 1,000,000 per year and plant. Hence, the accident risk is greatly reduced. The information given by the author are to smooth the emotional edge in the discussion about the safety of nuclear power plants. (orig.) [de

  11. Knowledge-Based operation planning system for boiling water reactors

    International Nuclear Information System (INIS)

    Tatsuya Iwamoto; Shungo Sakurai; Hitoshi Uematsu; Makoto Tsuiki

    1987-01-01

    A knowledge-Based Boiling Water Reactor operation planning system was developed to support core operators or core management engineers in making core operation plans, by automatically generating suboptimum core operation procedures. The procedures are obtained by searching a branching tree of the possible core status (nodes) and the elementary operations to change the core status (branches). A path that ends at the target node, and contains only operationally feasible nodes can be a candidate of the solution. The core eigenvalue, the power distribution and the thermal limit parameters at key points are calculated by running a three-dimensional (3-D) BWR core physics simulator to examine the feasibility of the nodes and the performance of candidates. To obtain a practically acceptable solution within a reasonable time rather than making a time-consuming effort to get the optimum one, the Depth-First-Search method, together with the heuristic branch-bounding, was used to search the branching tree. The system was applied to actual operation plannings with real plant data, and gave satisfactory results. It can be concluded that the system can be applied to generate core operation procedures as a substitute for core management experts

  12. Efficient characterization of fuel depletion in boiling water reactor

    International Nuclear Information System (INIS)

    Kim, S.H.

    1980-01-01

    An efficient fuel depletion method for boiling water reactor (BWR) fuel assemblies has been developed for fuel cycle analysis. A computer program HISTORY based on this method was designed to carry out accurate and rapid fuel burnup calculation for the fuel assembly. It has been usefully employed to study the depletion characteristics of the fuel assemblies for the preparation of nodal code input data and the fuel management study. The adequacy and the effectiveness of the assessment of this method used in HISTORY were demonstrated by comparing HISTORY results with more detailed CASMO results. The computing cost of HISTORY typically has been less than one dollar for the fuel assembly-level depletion calculations over the full life of the assembly, in contrast to more than $1000 for CASMO. By combining CASMO and HISTORY, a large number of expensive CASMO calculations can be replaced by inexpensive HISTORY. For the depletion calculations via CASMO/HISTORY, CASMO calculations are required only for the reference conditions and just at the beginning of life for other cases such as changes in void fraction, control rod condition and temperature. The simple and inexpensive HISTORY is sufficienty accurate and fast to be used in conjunction with CASMO for fuel cycle analysis and some BWR design calculations

  13. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  14. Generic risk insights for General Electric boiling water reactors

    International Nuclear Information System (INIS)

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

    1991-05-01

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

  15. Water inventory management in condenser pool of boiling water reactor

    International Nuclear Information System (INIS)

    Gluntz, D.M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs

  16. Critical Power Response to Power Oscillations in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Farawila, Yousef M.; Pruitt, Douglas W.

    2003-01-01

    The response of the critical power ratio to boiling water reactor (BWR) power oscillations is essential to the methods and practice of mitigating the effects of unstable density waves. Previous methods for calculating generic critical power response utilized direct time-domain simulations of unstable reactors. In this paper, advances in understanding the nature of the BWR oscillations and critical power phenomena are combined to develop a new method for calculating the critical power response. As the constraint of the reactor state - being at or slightly beyond the instability threshold - is removed, the new method allows the calculation of sensitivities to different operation and design parameters separately, and thus allows tighter safety margins to be used. The sensitivity to flow rate and the resulting oscillation frequency change are given special attention to evaluate the extension of the oscillation 'detect-and-suppress' methods to internal pump plants where the flow rate at natural circulation and oscillation frequency are much lower than jet pump plants

  17. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au)

  18. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

  19. The development of fuel elements for boiling water reactors

    International Nuclear Information System (INIS)

    Holzer, R.; Kilian, P.

    1984-01-01

    The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

  20. Introduction to nuclear power reactors and their health physics systems

    International Nuclear Information System (INIS)

    Brtis, J.S.

    1982-01-01

    This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

  1. State of nuclear waste management of German nuclear power stations

    International Nuclear Information System (INIS)

    1983-01-01

    The waste management of nuclear power plants in the Federal Republic of Germany is today prevailing in the public discussion. Objections raised in this connection, e.g. that the nuclear waste management has been omitted from the development of peaceful utilization of nuclear energy or remained insolved, are frequently accepted without examination, and partly spread as facts. This is, however, not the truth: From the outset in 1955 the development of nuclear technology in the Federal Republic of Germany has included investigations of the problems of reprocessing and non-detrimental disposal of radioactive products, and the results have been compiled in a national nuclear waste management concept. (orig.) [de

  2. Technical report on operating experience with boiling water reactor offgas systems

    International Nuclear Information System (INIS)

    Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

    1978-03-01

    Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

  3. 1984 availability of the world's nuclear power plants

    International Nuclear Information System (INIS)

    Szeless, A.; Oszuszky, F.

    1985-01-01

    This survey of the availability of the world's nuclear power plants in 1984 coveres 250 units (the CMEA countries excluded) with an aggregate 184,500 MWe, which are arranged by types of reactor and geographic distribution. The utilization of nuclear power plant capacity attained an average of 66% in 1984, which is an increase by 3 percentage points in the utilization of capacity over the previous year's level (63%). Capacity utilization in pressurized water reactors 1984 (69%) was 4 percentage points higher than it was in boiling water reactors (65%). The ranking list of the world's nuclear power plants is headed by one heavy water reactor and one gas cooled reactor each (98%), followed by five generating units, i.e., two pressurized water reactors, two boiling water reactors, and one heavy water reactor (all 95%). The best German nuclear power plants were Grafenrheinfeld (89%, position No. 20) and Unterweser and Stade (88%, position No. 25). (orig.) [de

  4. West German nuclear politics: a study of international cooperative behavior

    International Nuclear Information System (INIS)

    Driscoll, R.F.

    1983-01-01

    This dissertation is an investigation of the motivations behind West German policies and behavior in the area of nuclear politics. It examines and attempts to explain this behavior using a taxonomy built around the factors of vulnerability, power, domestic support, and autonomy. These are called international cooperative behavior factors. Chapters I and II develop the taxonomy and define the factors in terms of the Federal Republic. Chapter III covers the period up through the functioning of the Nuclear Planning Group, while Chapter IV analyzes NATO's December 1979 modernization decision. The presentation is less chronological and more selective: it focuses on the four international cooperative behavior factors as explanatory concepts. Chapter V examines the utility of the taxonomy. It concludes that the factors of autonomy, domestic support and vulnerability appear to be key in understanding and predicting West German nuclear behavior. The chapter then studies the potential applicability of the taxonomy to other nations or issues. It concludes that the factors are very nation-specific, but they do in fact provide a useful classification and explanatory scheme

  5. Activity buildup in the recirculation system of boiling water reactors

    International Nuclear Information System (INIS)

    Grauer, R.

    1987-05-01

    The deposition of activated corrosion products in the recirculation system of boiling water reactors produces increased radiation levels which lead to a corresponding increase in personnel irradiation dose rates during shut down and maintenance. The major part of this dose rate is due to Co-60. Based on a comprehensive literature study concerning this theme, the author has attempted to identify the individual stages of the activity build up and to classify their importance. The following areas are discussed in detail: - the origins of the corrosion products and of Co-59 in the reactor feedwater; - the consolidation of the cobalt in the fuel element crud deposits (activation); - the release and transport of the Co-60; - the build up of Co-60 in the corrosion products of the primary circuit. Existing models for the build up of cirquit radioactivity are discussed and the operating experience from selected reactors are summarized. Finally the state of the art of knowledge concerning the individual stages in the development of the activation build up is depicted. This highlights the existing gaps and thus identifies areas for possible R+D activities. Corrosion chemistry aspects of the cobald build up in the primary cirquit have already been studied on a broad basis and are continuing to be researched in a number of centers. The crystal chemistry of austenitic steel corrosion products poses a number of yet unanswered questions. There are major loopholes associated with an understanding of the activation processes of the cobalt deposited on the fuel elements and in the mass transfer of Co-60. For these processes, the most important influence stems from factors associated with colloid chemistry. In this respect there is ample room for new and original research contributions. It is recommended that sections 8 and 9 of this report should be read as an 'Executive Summary'. (author)

  6. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    Tsuiki, Makoto; Sakurada, Koichi; Yoshida, Hiroyuki.

    1988-01-01

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  7. Experimental study on a new solar boiling water system with holistic track solar funnel concentrator

    International Nuclear Information System (INIS)

    Xiaodi, Xue; Hongfei, Zheng; Kaiyan, He; Zhili, Chen; Tao, Tao; Guo, Xie

    2010-01-01

    A new solar boiling water system with conventional vacuum-tube solar collector as primary heater and the holistic solar funnel concentrator as secondary heater had been designed. In this paper, the system was measured out door and its performance was analyzed. The configuration and operation principle of the system are described. Variations of the boiled water yield, the temperature of the stove and the solar irradiance with local time have been measured. Main factors affecting the system performance have been analyzed. The experimental results indicate that the system produced large amount of boiled water. And the performance of the system has been found closely related to the solar radiance. When the solar radiance is above 600 W/m 2 , the boiled water yield rate of the system has reached 20 kg/h and its total energy efficiency has exceeded 40%.

  8. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  9. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  10. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  11. Numerical simulation of progressive inlet orifices in boiling water reactor fuel

    International Nuclear Information System (INIS)

    Lundgren, Sara

    2004-07-01

    This thesis was carried out at Forsmark Nuclear Power Plant. The power plant in Forsmark consists of three boiling water reactors (BWR) which produce about 17% of Swedish electricity. In a BWR the nuclear reactions are used to boil water inside the reactor vessel. The water works both as a coolant and as a moderator and the resulting steam is used directly to run the turbines. A problem when running a BWR at low flow conditions is the density wave oscillations that might occur to the water flow inside the fuel assemblies. These oscillations arise due to the connection between power and flow rate in a heated channel with two-phase flow. In order to improve the stability performance of the channel an orifice plate is placed at the inlet of each fuel assembly. Today these orifice plates have sharp edges and a constant resistance coefficient. Experimental work has been done with progressive orifices, the edge of which is half-oval in shape. The advantage of progressive orifices is the lower pressure losses with an increase of the Reynolds number, a similar phenomenon that appears in external flow around curved bodies. Since there are high costs associated with experimental generation of high- temperature and high-pressure data, it is of some interest to be able to reproduce and generate data using Computational Fluid Dynamics (CFD). This work deals with the possibility to use the CFD-code Fluent to do numerical simulations of the flow through progressive orifices. The following conclusions may be drawn from the numerical results: All simulations using Reynolds-Averaged Navier-Stokes (RANS) turbulence models, two-dimensional and three-dimensional, capture an abrupt decrease of the resistance coefficient at higher Reynolds numbers. Two-equation models seem to under-predict the critical Reynolds number. The five-equation Reynolds Stress Model (RSM) gives a critical Reynolds number of the same order of magnitude of that measured in experiments. No major differences have

  12. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  13. Development of a MELCOR self-initialization algorithm for boiling water reactors

    International Nuclear Information System (INIS)

    Chien, C.S.; Wang, S.J.; Cheng, S.K.

    1996-01-01

    The MELCOR code, developed by Sandia National Laboratories, is suitable for calculating source terms and simulating severe accident phenomena of nuclear power plants. Prior to simulating a severe accident transient with MELCOR, the initial steady-state conditions must be generated in advance. The current MELCOR users' manuals do not provide a self-initialization procedure; this is the reason users have to adjust the initial conditions by themselves through a trial-and-error approach. A MELCOR self-initialization algorithm for boiling water reactor plants has been developed, which eliminates the tedious trial-and-error procedures and improves the simulation accuracy. This algorithm adjusts the important plant variable such as the dome pressure, downcomer level, and core flow rate to the desired conditions automatically. It is implemented through input with control functions provided in MELCOR. The reactor power and feedwater temperature are fed as input data. The initialization work of full-power conditions of the Kuosheng nuclear power station is cited as an example. These initial conditions are generated successfully with the developed algorithm. The generated initial conditions can be stored in a restart file and used for transient analysis. The methodology in this study improves the accuracy and consistency of transient calculations. Meanwhile, the algorithm provides all MELCOR users an easy and correct method for establishing the initial conditions

  14. Evaluation of a severe accident management strategy for boiling water reactors -- Drywell flooding

    International Nuclear Information System (INIS)

    Yu, D.; Xing, L.; Kastenberg, W.E.; Okrent, D.

    1994-01-01

    Flooding of the drywell has been suggested as a strategy to prevent reactor vessel and containment failure in boiling water reactors. To evaluate the candidate strategy, this study considers accident management as a decision problem (''drywell flooding'' versus ''do nothing'') and develops a decision-oriented framework, namely, the influence diagram approach. This analysis chooses the long-term station blackout sequence for a Mark 1 nuclear power plant (Peach Bottom), and an influence diagram with a single decision node is constructed. The node probabilities in the influence diagram are obtained from US Nuclear Regulatory Commission reports or estimated by probabilistic risk assessment methodology. In assessing potential benefits compared with adverse effects, this analysis uses two consequence measures, i.e., early and late fatalities, as decision criteria. The analysis concludes that even though potential adverse effects exist, such as ex-vessel steam explosions and containment isolation failure, the drywell flooding strategy is preferred to ''do nothing'' when evaluated in terms of these consequence measures

  15. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    International Nuclear Information System (INIS)

    1994-01-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ''Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs

  16. German risk study on nuclear power stations. Phase B

    International Nuclear Information System (INIS)

    1989-11-01

    The German Risk Study on Nuclear Power Stations is concerned with investigations of accidents in nuclear facilities and their associated risks. These investigations are undertaken on behalf of the federal Minister of Research and Technology. They have been broken down into two phases (Phase A and Phase B). The results of Phase A were published in 1979 (GRS 79). This report contains a summary of the investigations relating to Phase B. After an introduction setting out the basic principles and aim of the study, a general review will be given of the most important results. The course of the investigations and the results have already been published in a Technical Report (GRS 89). (author)

  17. The safety analysis report for nuclear power plants in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Kohler, H.A.G.

    1976-01-01

    On the basis of the standard format for drawing up safety analysis reports for stationary fission reactors - this format was published in 1959 by the competent Federal Ministry for Atomic Energy - a report is made on the format and scope of German safety analysis reports. The reasons for drawing up a standard safety analysis report for nuclear power plants with pressurized water reactor or boiling water reactor and a KWU draft are discussed. (orig./RW) [de

  18. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Ramirez G, C.; Chavez M, C.

    2012-10-01

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  19. Bayesian optimization analysis of containment-venting operation in a boiling water reactor severe accident

    International Nuclear Information System (INIS)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu

    2017-01-01

    Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the “black-box” code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents

  20. Bayesian optimization analysis of containment-venting operation in a boiling water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu [Nuclear Safety Research Center, Japan Atomic Energy Agency, Ibaraki (Japan)

    2017-03-15

    Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the “black-box” code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

  1. Remote mechanized equipment for the repair and replacement of boiling water reactor recirculation loop piping

    International Nuclear Information System (INIS)

    Mauser, D.; Busch, D.F.

    1983-01-01

    Equipment has been assembled for the remote repair or replacement of boiling water reactor nuclear plant piping in the diameter range of 4 to 28 inches (10-71 cm). The objectives of this program were to produce high-quality pipe welds, reduce plant downtime, and reduce man-rem exposure. The repair strategy was to permit repair personnel to install and check out the repair subsystems and then leave the radiation zone allowing the operations to be conducted at a distance of up to 300 feet (91 m) from the operator. The complete repair system comprises subsystems for pipe severing, dimensional gaging, joint preparation, counterboring, welding, postweld nondestructive inspection (conceptual design), and audio, electronic, and visual monitoring of all operations. Components for all subsystems, excluding those for postweld nondestructive inspection, were purchased and modified as needed for integration into the repair system. Subsystems were designed for two sizes of Type 304 stainless steelpipe. For smaller, 12-inch-diameter (30.5 cm) pipe, severing is accomplished by a power hack saw and joint preparation and counterboring by an internally mounted lathe. The 22-inch-diameter (56 cm) pipe is severed, prepared, and counterbored using an externally mounted, single-point machining device. Dimensional gaging is performed to characterize the pipe geometry relative to a fixed external reference surface, allowing the placement of the joint preparation and the counterbore to be optimized. For both pipe sizes, a track-mounted gas tungsten-arc welding head with filler wire feed is used

  2. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  3. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    International Nuclear Information System (INIS)

    Lanthen, Jonas

    2006-09-01

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes

  4. A proof-of-concept transient diagnostic expert system for BWRs [Boiling Water Reactors

    International Nuclear Information System (INIS)

    Yoshida, K.; Naser, J.A.

    1988-05-01

    A proof-of-concept transient diagnostic expert system has been developed to identify the cause and the type of an abnormal transient in a boiling water nuclear power plant. For this expert system development, the calculational results of the simulation code RETRAN were used as the knowledge source. The knowledge extracted from the RETRAN analyses was transformed into IF-THEN rules in the knowledge base for the expert system. An important feature of this expert system is the introduction of certainty factors to allow diagnosis even in the cases where data may be either missing or marked as invalid. To increase the capability of this diagnostic system to distinguish between similiar transients, backward chaining reasoning is used to support the forward chaining reasoning with certainty factors. Through this effort, it has been demonstrated that an expert system can be successfully used to create a transient diagnostic system. It has also successfully demonstrated that RETRAN can be used as the knowledge source for developing the knowledge base of the diagnostic system

  5. Automatic boiling water reactor loading pattern design using ant colony optimization algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C.-D. [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China); Nuclear Engineering Division, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)], E-mail: jdwang@iner.gov.tw; Lin Chaung [Department of Engineering and System Science, National Tsing Hua University, 101, Section 2 Kuang Fu Road, Hsinchu 30013, Taiwan (China)

    2009-08-15

    An automatic boiling water reactor (BWR) loading pattern (LP) design methodology was developed using the rank-based ant system (RAS), which is a variant of the ant colony optimization (ACO) algorithm. To reduce design complexity, only the fuel assemblies (FAs) of one eight-core positions were determined using the RAS algorithm, and then the corresponding FAs were loaded into the other parts of the core. Heuristic information was adopted to exclude the selection of the inappropriate FAs which will reduce search space, and thus, the computation time. When the LP was determined, Haling cycle length, beginning of cycle (BOC) shutdown margin (SDM), and Haling end of cycle (EOC) maximum fraction of limit for critical power ratio (MFLCPR) were calculated using SIMULATE-3 code, which were used to evaluate the LP for updating pheromone of RAS. The developed design methodology was demonstrated using FAs of a reference cycle of the BWR6 nuclear power plant. The results show that, the designed LP can be obtained within reasonable computation time, and has a longer cycle length than that of the original design.

  6. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  7. On-line test of power distribution prediction system for boiling water reactors

    International Nuclear Information System (INIS)

    Nishizawa, Y.; Kiguchi, T.; Kobayashi, S.; Takumi, K.; Tanaka, H.; Tsutsumi, R.; Yokomi, M.

    1982-01-01

    A power distribution prediction system for boiling water reactors has been developed and its on-line performance test has proceeded at an operating commercial reactor. This system predicts the power distribution or thermal margin in advance of control rod operations and core flow rate change. This system consists of an on-line computer system, an operator's console with a color cathode-ray tube, and plant data input devices. The main functions of this system are present power distribution monitoring, power distribution prediction, and power-up trajectory prediction. The calculation method is based on a simplified nuclear thermal-hydraulic calculation, which is combined with a method of model identification to the actual reactor core state. It has been ascertained by the on-line test that the predicted power distribution (readings of traversing in-core probe) agrees with the measured data within 6% root-mean-square. The computing time required for one prediction calculation step is less than or equal to 1.5 min by an HIDIC-80 on-line computer

  8. Thermal analysis and design of a passive reflux condenser for the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Bijlani, C.; Patti, F.; Prasad, V.

    1993-01-01

    At present, the advanced light water reactors (ALWRS) in the United States are being designed to remove reactor decay heat for a period of 72 h following a postulated loss-of-coolant accident (LOCA). The water in the pools external to the containment is evaporated or boiled off to remove the decay heat. It is presumed that the water in the pools can be replenished within 72 h through operator actions or outside assistance. Some countries in Europe require that the plant be designed to remove the reactor decay heat for a much longer duration than 72 h without external assistance. This paper presents an analysis and design of a passive heat exchanger called a reflux condenser (RC), which was considered for an ALWR-the 600-MW(electric) simplified boiling water reactor. The RC is required to condense the steam formed when the water in the pool in which the passive containment cooling system (PCCS) is immersed boils following a LOCA. The RCs are nuclear non-safety related. This paper presents steady-state performance of an RC at various outdoor air dry-bulb temperatures under still air conditions

  9. Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

    International Nuclear Information System (INIS)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

  10. Studies on improvements in the control methods of boiling water reactor plant

    International Nuclear Information System (INIS)

    Mankin, Shuichi

    1982-08-01

    In order to improve the performance of regulation and load following control of boiling water reactor plant, optimal control theory is applied and new types of control method are developed. Case-α controller is first formulated on the basis of the optimal linear regulator theory applied to the linealized model of the system; it is then modified by adding a integration-type action in a feed back loop and by the use of variable gain and reference for adapting to the power level requested. Case-#betta# controller consists of a hierarchical control scheme which has classical P.I. type sub-loop controllers at the first level and a linear optimal regulator at the second level. The controller is designed on the basis of the optimal regulator theory applied to the multivariate autoregressive system model which is obtained from the identification experiments, where the system model is determined with the conventional sub-loop controllers included. The results of the simulation experiments show these control methods proposed have performed fairly well and will be useful for the improvement of the performance of nuclear power plant control. In addition, it is suggested that these control methods will be also attractive for the control of other production plants because these were developed in the attempt to solve the problems deviated from so called 'The gap between the optimal contro theory and actual systems.' (author)

  11. Operating data monitoring and fatigue evaluation systems and findings for boiling water reactors in Japan

    International Nuclear Information System (INIS)

    Maekawa, O.; Kanazawa, Y.; Takahashi, Y.; Tani, M.

    1995-01-01

    The reactor pressure vessel (RPV) is one of the most critical components of a boiling water reactor (BWR) when utilities think about plant life extension (PLEX). Design stress analysis sometimes reports very high fatigue usage factors for such portions of RPVs as stud bolt, feedwater nozzle and support skirt.In order to evaluate design margin and to eliminate excessive conservatism in this design analysis to pave the way for PLEX, Japanese BWR utilities jointly with BWR manufacturers in Japan established a programme (1) to acquire plant operational data on line for specific parameters used in stress analysis, (2) to evaluate margin in the design using measured plant data best estimate boundary conditions for stress analysis, and (3) to establish a simplified fatigue analysis method for BWR RPV.A plant data acquisition system, named OPEDAS, has been developed and installed in Tokyo Electric Power Company's 1100MWe BWR at its Kashiwazaki-Kariwa nuclear power plant. Best estimate stress analysis using measured in-plant data has been carried out and the results show considerable margin in fatigue usage factor over the design. A simplified fatigue analysis method using in-plant data has been developed with the Green's function, although some limitations have been identified for its use. ((orig.))

  12. R&D Activities in German Nuclear Sector

    International Nuclear Information System (INIS)

    Schwenk-Ferrero, A.

    2015-01-01

    Germany is phasing out its nuclear energy use. The seven oldest Light Water Reactors (taken offline in 2011 for a safety review immediately following the Fukushima accident) and the Kruemmel Nuclear Power Plant (NPP) are already shut down; six others would go offline by 2021 at the latest and the three newest by 2022. Phasing out nuclear power must not be regarded as synonymous with “phasing out” nuclear expertise. Long after Germany has completed its NPP shutdown, such skills will remain essential for activities such as ensuring reactor safety, radiation protection, decommissioning, ultimate disposal of radioactive waste and crisis management, and for maintaining a critical outlook on international developments. The German Government underlined this fact when launching the 6th Energy Research Programme entitled “Research for an Environmentally Sound, Reliable and Affordable Energy Supply” in August 2011. The dispute on questions of nuclear waste management and disposal in Germany is not only framed by the political parties’ different perspectives and conflicts between the antinuclear movement and official politics, but is also characterised by the dissent among experts regarding fundamental questions of nuclear waste storage. Especially for high-level radioactive waste there are still debates about the possible repository type and site. In 2014 the study on “Partitioning and Transmutation of Nuclear Waste: benefits and Challenges in Research and Implementation” has been performed under the supervision of National Academy of Science and Engineering–acatech. Acatech recommended continuing the R&D activities on P&T in Germany within a wide European framework and examining the options for future participation of Germany in P&T oriented research in Europe in order to follow a European perspective with a possible benefit to German nuclear industry. German research should concentrate on the key areas: efficient partitioning and transmutation of

  13. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  14. Filters in nuclear facilities

    International Nuclear Information System (INIS)

    Berg, K.H.; Wilhelm, J.G.

    1985-01-01

    The topics of the nine papers given include the behavior of HEPA filters during exposure to air flows of high humidity as well as of high differential pressure, the development of steel-fiber filters suitable for extreme operating conditions, and the occurrence of various radioactive iodine species in the exhaust air from boiling water reactors. In an introductory presentation the German view of the performance requirements to be met by filters in nuclear facilities as well as the present status of filter quality assurance are discussed. (orig.) [de

  15. German study on the risks of nuclear power plants

    International Nuclear Information System (INIS)

    Wollny, V.

    1987-01-01

    The 'Deutsche Risikostudie Kernkraftwerke', DRS, (German study on the risks of nuclear plants) calculates a frequency of 1/10000 years (i.e. once in 10000 years of reactor operation) for the event of insufficient cooling of the reactor core occurring. The resulting collective risk is put at approximately 10 deaths per year for 25 plants. As the explanations show, the error margin of such a risk statement is enormous. Moreover, much seems to indicate that the risk was underrated. Apart from this, risk analyses are of doubtful value in the decision process for or against the introduction of large-scale technologies. They cannot replace the individual decision of all concerned. (orig./HSCH) [de

  16. West German alternatives for reducing reliance on nuclear weapons

    International Nuclear Information System (INIS)

    Brauch, H.G.

    1988-01-01

    Nuclear deterrence in general and the US doctrinal concept of mutual assured destruction as well as the common strategic understanding of mutual vulnerability, codified by the SALT process, and the nuclear first use option of NATO's flexible response strategy (MC14/3) have come under attack, both from official and unofficial circles, in the United States and Europe likewise. In spite of the general agreement, that the time may have come to search for alternatives, to move from MAD to MAS (mutual assured security), as President Reagan indicated in a New York Times interview or for a Common Security posture, as called for by the Palme Commission, nevertheless major disagreements and contradictions exist as to how European security could be enhanced with a defensive posture beyond deterrence. Two groups of alternatives are being distinguished in this paper: official efforts in the United States and in Europe, to use the anti-nuclear sentiment, to legitimate changes in the operative doctrine of the US and of NATO forces and to obtain funds both for a comprehensive buildup and modernization of conventional forces in Europe and for a Strategic (SDI) and probably soon also for a European Defence Initiative (EDI) or an extended air defence, and unofficial proposals by American, European and West German Experts for a non-provocative, inoffensive defence or for a gradual defensivity consisting of static and mobile components

  17. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  18. Coarse mesh finite element method for boiling water reactor physics analysis

    International Nuclear Information System (INIS)

    Ellison, P.G.

    1983-01-01

    A coarse mesh method is formulated for the solution of Boiling Water Reactor physics problems using two group diffusion theory. No fuel assembly cross-section homogenization is required; water gaps, control blades and fuel pins of varying enrichments are treated explicitly. The method combines constrained finite element discretization with infinite lattice super cell trial functions to obtain coarse mesh solutions for which the only approximations are along the boundaries between fuel assemblies. The method is applied to bench mark Boiling Water Reactor problems to obtain both the eigenvalue and detailed flux distributions. The solutions to these problems indicate the method is useful in predicting detailed power distributions and eigenvalues for Boiling Water Reactor physics problems

  19. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-04-30

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

  20. Disposal of spent fuel from German nuclear power plants - 16028

    International Nuclear Information System (INIS)

    Graf, Reinhold; Brammer, Klaus-Juergen; Filbert, Wolfgang; Bollingerfehr, Wilhelm

    2009-01-01

    The 'direct disposal of spent fuel' as a part of the current German reference concept was developed as an alternative to spent fuel reprocessing and vitrified HLW disposal. The technical facilities necessary for the implementation of this part of the reference concept, the so called POLLUX R concept, i.e. interim storage buildings for casks containing spent fuel, a pilot conditioning facility, and a special cask 'POLLUX' for final disposal have been built. With view to a geological salt formation all handling procedures for the direct disposal of spent fuel were tested aboveground in full-scale test facilities. To optimise the reference concept, all operational steps have been reviewed for possible improvements. The two additional concepts for the direct disposal of SF are the BSK 3 concept and the DIREGT concept. Both concepts rely on borehole emplacement technology, vertical boreholes for the BSK 3 concept und horizontal boreholes for the DIREGT concept. Supported by the EU and the German Federal Ministry of Economics and Technology (BMWi), DBE TECHNOLOGY built an aboveground full-scale test facility to simulate all relevant handling procedures for the BSK 3 disposal concept. GNS (Company for Nuclear Service), representing the German utilities, provided the main components and its know-how concerning cask design and manufacturing. The test program was concluded recently after more than 1.000 emplacement operations had been performed successfully. The BSK 3 emplacement system in total comprises an emplacement device, a borehole lock, a transport cart, a transfer cask which will shuttle between the aboveground conditioning facility and the underground repository, and the BSK 3 canister itself, designed to contain the fuel rods of three PWR-fuel assemblies with a total of about 1.6 tHM. The BSK 3 concept simplifies the operation of the repository because the handling procedures and techniques can also be applied for the disposal of reprocessing residues. In addition

  1. Hitler's bomb: the secret story of Germans' attempts to get the nuclear weapon

    International Nuclear Information System (INIS)

    Karlsch, Rainer

    2007-01-01

    In this historical book, the author claims to have evidence concerning the development and testing of a possible 'nuclear weapon' by Nazi Germany in 1945. The 'weapon' in question is not alleged to be a standard nuclear weapon powered by nuclear fission, but something closer to either a radiological weapon (a so-called 'dirty bomb') or a hybrid-nuclear fusion weapon. Its new evidence is concerned primarily with the parts of the German nuclear energy project (an attempted clandestine scientific effort led by Germany to develop and produce atomic weapons during World War II) under Kurt Diebner, a German nuclear physicist who directed and administrated the project

  2. Basic philosophy of the safety design of the Toshiba boiling water reactor

    International Nuclear Information System (INIS)

    Sato, T.

    1992-01-01

    This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

  3. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

  4. Numerical evaluation of fluid mixing phenomena in boiling water reactor using advanced interface tracking method

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Takase, Kazuyuki

    2008-01-01

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low. (author)

  5. Cryogenic system for collecting noble gases from boiling water reactor off-gas

    International Nuclear Information System (INIS)

    Schmauch, G.E.

    1973-01-01

    In boiling water reactors, noncondensible gases are expelled from the main condenser. This off-gas stream is composed largely of radiolytic hydrogen and oxygen, air in-leakage, and traces of fission product krypton and xenon. In the Air Products' treatment system, the stoichiometric hydrogen and oxygen are reacted to form water in a catalytic recombiner. The design of the catalytic recombiner is an extension of industrial gas technology developed for purification of argon and helium. The off-gas after the recombiner is processed by cryogenic air-separation technology. The gas is compressed, passed into a reversing heat exchanger where water vapor and carbon dioxide are frozen out, further cooled, and expanded into a distillation column where refrigeration is provided by addition of liquid nitrogen. More than 99.99 percent of the krypton and essentially 100 percent of the xenon entering the column are accumulated in the column bottoms. Every three to six months, the noble-gas concentrate accumulated in the column bottom is removed as liquid, vaporized, diluted with steam, mixed with hydrogen in slight excess of oxygen content, and fed to a small recombiner where all the oxygen reacts to form water. The resulting gas stream, containing from 20 to 40 percent noble gases, is compressed into small storage cylinders for indefinite retention or for decay of all fission gases except krypton-85, followed by subsequent release under controlled conditions and favorable meteorology. This treatment system is based on proven technology that is practiced throughout the industrial gas industry. Only the presence of radioactive materials in the process stream and the application in a nuclear power plant environment are new. Adaptations to meet these new conditions can be made without sacrificing performance, reliability, or safety

  6. Bifurcation analysis of the simplified models of boiling water reactor and identification of global stability boundary

    Energy Technology Data Exchange (ETDEWEB)

    Pandey, Vikas; Singh, Suneet, E-mail: suneet.singh@iitb.ac.in

    2017-04-15

    Highlights: • Non-linear stability analysis of nuclear reactor is carried out. • Global and local stability boundaries are drawn in the parameter space. • Globally stable, bi-stable, and unstable regions have been demarcated. • The identification of the regions is verified by numerical simulations. - Abstract: Nonlinear stability study of the neutron coupled thermal hydraulics instability has been carried out by several researchers for boiling water reactors (BWRs). The focus of these studies has been to identify subcritical and supercritical Hopf bifurcations. Supercritical Hopf bifurcation are soft or safe due to the fact that stable limit cycles arise in linearly unstable region; linear and global stability boundaries are same for this bifurcation. It is well known that the subcritical bifurcations can be considered as hard or dangerous due to the fact that unstable limit cycles (nonlinear phenomena) exist in the (linearly) stable region. The linear stability leads to a stable equilibrium in such regions, only for infinitesimally small perturbations. However, finite perturbations lead to instability due to the presence of unstable limit cycles. Therefore, it is evident that the linear stability analysis is not sufficient to understand the exact stability characteristics of BWRs. However, the effect of these bifurcations on the stability boundaries has been rarely discussed. In the present work, the identification of global stability boundary is demonstrated using simplified models. Here, five different models with different thermal hydraulics feedback have been investigated. In comparison to the earlier works, current models also include the impact of adding the rate of change in temperature on void reactivity as well as effect of void reactivity on rate of change of temperature. Using the bifurcation analysis of these models the globally stable region in the parameter space has been identified. The globally stable region has only stable solutions and

  7. Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M

    1962-05-15

    This paper deals with a new concept for predicting burnout conditions for forced convection of boiling water in fuel elements of nuclear boiling reactors. The concept states the importance of considering the ratio of heated channel perimeter to total channel perimeter. The perimeter ratio concept was arrived at from an experimental study of burnout conditions in rod clusters consisting of three rods of 13 mm outside diameter and 970 mm heated length. Data were obtained for pressures between{sub 2}. 5 and 10 kg/cm, surface heat fluxes between 50 and 120 W/cm, mass flow rates between 0.03 and 0.33 kg/sec and steam qualities between 0.01 and 0.52. The rod distances for the experiment were 2 mm and 6 mm. The diameter of the channel was 41.3 mm. Additional runs were also performed after introducing unheated displacement rods in the channel. The rod distance in this case was 6 mm. In the ranges investigated the measured burnout steam qualities at the outlet of the channel decreases with increasing heat flux and decreasing pressure. Furthermore it has been found that the influence of rod distance is, in the range investigated, of small significance for engineering purposes. It has also been observed that the present burnout steam quality data for the rod clusters are much lower than those earlier obtained for round ducts. This may be explained physically by means of the perimeter ratio concept. It has also been found that the surface shear-stress distribution around the channel perimeter and especially the position of maximum shear-stress is of great importance for predicting burnout conditions for flow in channels. Finally the new method has helped us to understand and interpret experimental results which earlier may have seemed inconsistent.

  8. Bifurcation analysis of the simplified models of boiling water reactor and identification of global stability boundary

    International Nuclear Information System (INIS)

    Pandey, Vikas; Singh, Suneet

    2017-01-01

    Highlights: • Non-linear stability analysis of nuclear reactor is carried out. • Global and local stability boundaries are drawn in the parameter space. • Globally stable, bi-stable, and unstable regions have been demarcated. • The identification of the regions is verified by numerical simulations. - Abstract: Nonlinear stability study of the neutron coupled thermal hydraulics instability has been carried out by several researchers for boiling water reactors (BWRs). The focus of these studies has been to identify subcritical and supercritical Hopf bifurcations. Supercritical Hopf bifurcation are soft or safe due to the fact that stable limit cycles arise in linearly unstable region; linear and global stability boundaries are same for this bifurcation. It is well known that the subcritical bifurcations can be considered as hard or dangerous due to the fact that unstable limit cycles (nonlinear phenomena) exist in the (linearly) stable region. The linear stability leads to a stable equilibrium in such regions, only for infinitesimally small perturbations. However, finite perturbations lead to instability due to the presence of unstable limit cycles. Therefore, it is evident that the linear stability analysis is not sufficient to understand the exact stability characteristics of BWRs. However, the effect of these bifurcations on the stability boundaries has been rarely discussed. In the present work, the identification of global stability boundary is demonstrated using simplified models. Here, five different models with different thermal hydraulics feedback have been investigated. In comparison to the earlier works, current models also include the impact of adding the rate of change in temperature on void reactivity as well as effect of void reactivity on rate of change of temperature. Using the bifurcation analysis of these models the globally stable region in the parameter space has been identified. The globally stable region has only stable solutions and

  9. Channel Bow in Boiling Water Reactors - Hot Cell Examination Results and Correlation to Measured Bow

    International Nuclear Information System (INIS)

    Mahmood, S.T.; Lin, Y.P.; Dubecky, M.A.; Edsinger, K.; Mader, E.V.

    2007-01-01

    An increase in frequency of fuel channel-control blade interference has been observed in Boiling Water Reactors (BWR) in recent years. Many of the channels leading to interference were found to bow towards the control blade in a manner that was inconsistent with the expected bow due to other effects. The pattern of bow appeared to indicate a new channel bow mechanism that differed from the predominant bow mechanism caused by differential growth due to fast-fluence gradients. In order to investigate this new type of channel bow, coupons from several channels with varying degrees of bow were returned to the GE Vallecitos Nuclear Center (VNC) for Post-Irradiation Examination (PIE). This paper describes the characteristics of channel corrosion and hydrogen pickup observed, and relates the observations to the channel exposure level, control history, and measured channel bow. The channels selected for PIE had exposures in the range of 36-48 GWd/MTU and covered a wide range of measured bow. The coupons were obtained at 4 elevations from opposing channel sides adjacent and away from the control blade. The PIE performed on these coupons included visual examination, metallography, and hydrogen concentration measurements. A new mechanism of control-blade shadow corrosion-induced channel bow was found to correlate with differences in the extent of corrosion and corresponding differences in the hydrogen concentration between opposite sides of the channels. The increased corrosion on the control blade sides was found to be dependent on the level of control early in the life of the channel. The contributions of other potential factors leading to increased channel bow and channel-control blade interference are also discussed in this paper. (authors)

  10. Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

    International Nuclear Information System (INIS)

    Becker, Kurt M.

    1962-05-01

    This paper deals with a new concept for predicting burnout conditions for forced convection of boiling water in fuel elements of nuclear boiling reactors. The concept states the importance of considering the ratio of heated channel perimeter to total channel perimeter. The perimeter ratio concept was arrived at from an experimental study of burnout conditions in rod clusters consisting of three rods of 13 mm outside diameter and 970 mm heated length. Data were obtained for pressures between 2 . 5 and 10 kg/cm, surface heat fluxes between 50 and 120 W/cm, mass flow rates between 0.03 and 0.33 kg/sec and steam qualities between 0.01 and 0.52. The rod distances for the experiment were 2 mm and 6 mm. The diameter of the channel was 41.3 mm. Additional runs were also performed after introducing unheated displacement rods in the channel. The rod distance in this case was 6 mm. In the ranges investigated the measured burnout steam qualities at the outlet of the channel decreases with increasing heat flux and decreasing pressure. Furthermore it has been found that the influence of rod distance is, in the range investigated, of small significance for engineering purposes. It has also been observed that the present burnout steam quality data for the rod clusters are much lower than those earlier obtained for round ducts. This may be explained physically by means of the perimeter ratio concept. It has also been found that the surface shear-stress distribution around the channel perimeter and especially the position of maximum shear-stress is of great importance for predicting burnout conditions for flow in channels. Finally the new method has helped us to understand and interpret experimental results which earlier may have seemed inconsistent

  11. Three-Dimensional Modeling of a Steam-Line Break in a Boiling Water Reactor

    International Nuclear Information System (INIS)

    Tinoco, Hernan

    2002-01-01

    Because of weld problems, the core grids of Units 1 and 2 at the Forsmark nuclear power plant have been replaced by grids of a new design, consisting of a single machined piece without welds. The qualifying structural analysis has been carried out considering dynamic loads, which implies that even loss-of-coolant accidents have to be included. Therefore, a detailed time description of the loads acting on the different internal parts of the reactor is needed. To achieve sufficient space and time resolution, a computational fluid dynamics (CFD) analysis was considered to be a viable alternative.A CFD analysis of a steam-line break in the boiling water reactor of Unit 2 is the subject of this work. The study is based on the assumption that the timescale of the transient analysis is smaller than the relaxation time of the water-steam system. Therefore, a simulation of only the upper, steam part of the reactor with no two-phase effects (flashing) is feasible.The results obtained display a rather complex behavior of the decompression process, forcing the analysis of the pressure field to be accomplished through animation. In contrast, the computed instantaneous forces over different internal parts oscillate regularly and are approximately twice the forces estimated in the past by simpler methods, with frequencies of 30 to 40 Hz; top amplitudes of ∼1.64 MN; and relatively low damping, ∼25% after 0.5 s.According to the present results, this type of modeling is physically meaningful for simulation timescales smaller than the water-steam relaxation time, i.e., ∼0.5 s at reactor conditions. At larger times, a two-phase model is necessary to describe the decompression process since two-phase effects are dominant. The results have not yet been validated with experiments, but validation computations will be run in the future for comparison with results of the Marviken tests

  12. A real-time BWR [boiling water reactor] stability measurement system

    International Nuclear Information System (INIS)

    March-Leuba, J.; King, W.T.

    1987-01-01

    This paper describes the characteristics of a portable, real-time system used for nonperturbational measurements of stability in boiling water reactors. The algorithm used in this system estimates the closed-loop asymptotic decay ratio using only the naturally occurring neutron noise and it is based on the univariate autoregressive methodology

  13. Instrumenting a pressure suppression experiment for a MK I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1977-01-01

    A scale test facility of a pressure suppression system from a boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy experimental data during a hypothetical loss-of-coolant accident. The instrumentation verified the analysis of the dynamic loading of the pressure suppression system

  14. Generation of shockwave and vortex structures at the outflow of a boiling water jet

    Science.gov (United States)

    Alekseev, M. V.; Lezhnin, S. I.; Pribaturin, N. A.; Sorokin, A. L.

    2014-12-01

    Results of numerical simulation for shock waves and generation of vortex structures during unsteady outflow of boiling liquid jet are presented. The features of evolution of shock waves and vortex structures formation during unsteady outflow of boiling water are compared with corresponding structures during unsteady gas outflow.

  15. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  16. Generic safety insights for inspection of boiling water reactors

    International Nuclear Information System (INIS)

    Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

    1987-01-01

    As the number of operating nuclear power plants (NPP) increases, safety inspection has increased in importance. However, precisely what is important, and what is not important? What should one focus inspection efforts on. Over the last two years Probabilistic Risk Assessment (PR) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only about 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the BNL program to develop generic BWR PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs

  17. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    International Nuclear Information System (INIS)

    Ishii, M.; Revankar, S. T.; Downar, T.; Xu, Y.; Yoon, H. J.; Tinkler, D.; Rohatgi, U. S.

    2003-01-01

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  18. First domestic primary loop recircuration pump for boiling water reactor

    International Nuclear Information System (INIS)

    Fukuda, Minoru; Taka, Shusei; Kato, Hiroyuki

    1981-01-01

    Two primary loop recirculation (PLR) pumps for the second unit of the Fukushima No. 2 Nuclear Power Station of the Tokyo Electric Power Co., Inc., have been manufactured by Ebara Corporation. They are the first domestically produced pumps for commercial power plants and were manufactured under license from Byron Jackson Pump Division of Borg Warner Corporation. This article describes the special features of pump design and stress analysis, and the results of the 700 hours of factory loop tests, which are all essential for the PLR pump. (author)

  19. Main results of the German nuclear power plant risk study

    International Nuclear Information System (INIS)

    Danzmann, H.J.

    1981-01-01

    The subject is discussed under the headings: introduction; purpose and task of the German risk study; approach; results of investigations (analyses of engineered plant features; determination of accident consequences); emergency response model; protective actions and countermeasures; evaluation. (U.K.)

  20. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mkrtchyan, Lilit; Schau, Henry [TUEV SUED Energietechnik GmbH, Mannheim (Germany). Abt. Strukturverhalten; Wolf, Werner; Holzer, Wieland [TUEV SUED Industrie Service GmbH, Muenchen (Germany). Abt. Behaelter und Turbosatz; Wernicke, Robert; Trieglaff, Ralf [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany). Abt. Festigkeit und Konstruktion

    2011-08-15

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  1. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    International Nuclear Information System (INIS)

    Mkrtchyan, Lilit; Schau, Henry; Wolf, Werner; Holzer, Wieland; Wernicke, Robert; Trieglaff, Ralf

    2011-01-01

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  2. Recent experiences with ultrasonic inservice inspection systems with phased array probes on spherical bottoms of boiling water reactors

    International Nuclear Information System (INIS)

    Wustenberg, H.; Brekow, G.; Erhard, A.; Hein, E.

    1988-01-01

    The special geometry of the spherical bottom of boiling water reactors with control rods and measuring nozzles requires a very special surveillance technique during the in-service inspection. Reside visual inspection an ultrasonic inspection has been established due to the requirements of German authorities. A first application of a new phased array system took place August 1987. The 100% inspection of a spherical bottom had been enabled by the application of phased array probes with electronically controlled skewing angles. The data acquisition had been based on the storage of whole A-scans, which had been pixellized into 256 points. This A-scan storage procedure makes possible the application of a simple and fast algorithm to present the data as TD-(time displacement)-scans. Defect reconstruction by echotomographique approaches are under development. This paper presents the ultrasonic technique applied including the phased array probes, the electronic system, as well as the software package used for the control of the inspection parameters depending on the probe position

  3. Development of an in-core fuel management tool for boiling water reactors

    International Nuclear Information System (INIS)

    Gilli, Luca; Wakker, Pieter H.; Elder, Brian R.

    2017-01-01

    The in-core fuel management of a nuclear reactor is a challenging task due to the virtually infinite number of loading patterns one could theoretically adopt. The ROSA (Reloading Optimization by Simulated Annealing) code is an optimization tool that has been successfully used in the last two decades to facilitate the core design of several Pressurized Water Reactors (PWRs). It is designed to perform a stochastic search for an optimal Loading Pattern (LP) using a simulated annealing algorithm. This corresponds to performing a depletion calculation for each one of the hundreds of thousands of unique LPs generated during the stochastic search. Therefore, speed is one of the most important requirements that the solvers used by the depletion tool must fulfill. ROSA's depletion analysis tool makes use of a particularly fast nodal method (known as the kernel method) for the evaluation of the power distribution associated with a particular LP. One of the strongest assumptions behind the kernel method is that the neutron migration length does not change considerably between the point where a neutron is generated and the point where the same neutron is absorbed. Although strong, this assumption is quite compatible with the neutronic characteristics of PWRs cores. In this paper we give an overview of the work done in order to develop a version of ROSA capable of performing the core design of Boiling Water Reactors (BWRs). We focus the discussion on the development of the depletion analysis tool by outlining the modifications of the kernel methods implemented in order to make the solver accurate for BWR cores. An improvement of the definition of the transport kernel is necessary to take the strong anisotropies characterizing the neutronic problem into account. These anisotropies arise due to the presence of strong changes in the moderator density and due to the presence of control blades. Furthermore, we are going to discuss how the boundary conditions are adopted by the

  4. RELAP 4/MOD 6 boiling water nodalization study

    International Nuclear Information System (INIS)

    Sonneck, G.; Pfau, H.

    1985-09-01

    The risk of nuclear steam supply systems is dominated by the core melt accidents. The first step to a realistic assessment of these sequences is the successful prediction of a loss of coolant event in a test loop. One of the codes for that is RELAP 4/MOD 6 and one of the important options in this code is the nodalization. The base of this work is the test LOCA No. 1 FIX II in Studsvik (Sweden) which also served as the OECD International Standard Problem 15. This report discusses the influence of different nodalizations, of different distributions of pressure, water and structural heat as well as of different bubble rise options, break flow coefficients, and heat transfer time steps. The most important result is that a simple RELAP 4/MOD6 model with less than 10 volumes is able to predict an experiment as LOCA No. 1 in FIX II successfully using only a fraction of the usual computing time. (Author)

  5. Nuclear power in the energy economics of the German Democratic Republic

    International Nuclear Information System (INIS)

    Hildebrand, H.J.

    1977-01-01

    The present structure of the energy economy in the German Democratic Republic (East Germany) is described and the need of nuclear power stations for the future supply of energy is demonstrated. The selection of the type of nuclear reactor, the existing operating experience with nuclear equipment and the collaboration in this field with the U.S.S.R. is described. The training of personnel in nuclear techniques is noted. (H.E.G.)

  6. On-site storage of spent nuclear fuel assemblies in German nuclear power plants

    International Nuclear Information System (INIS)

    Banck, J.

    1999-01-01

    The selection of back-end strategies for spent fuel assemblies is influenced by a number of different factors depending on the given situation in any specific country. In Germany, the back-end strategy implemented in the past was almost exclusively reprocessing. This strategy was required by the German Atomic Energy Act. Since 1994, when the Atomic Energy Act was amended, the option of direct final disposal has been granted the equivalent status by law to that afforded to reprocessing (and reuse of valuable materials). As a result, German utilities may now choose between these two alternatives. Another important condition for optimizing the back-end policy is the fact that fuel cycle costs in Germany are directly dependent on spent fuel volumes (in contrast to the US, for example, such costs are related to the amount of power generated). Another boundary condition for German utilities with respect to spent fuel management is posed by the problems with militant opponents of nuclear energy during transportation of spent fuel to interim storage sites. These facts have given rise to a reconsideration of the fuel cycle back-end, which has resulted in a change in strategy by most German utilities in favour of the following: Preference for long-term storage and maximized use of on-site storage capacity; Reduction in the amount of spent fuel by increasing burnup as much as possible. These decisions have also been driven by the deregulation of energy markets in Europe, where utilities are now permitted to sell electric power to consumers beyond their original supply network and must therefore offer electric power on a very cost competitive basis. (author)

  7. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  8. Development of alarm handling methods for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yukiharu, Ohga; Hiroshi, Seki; Setsuo, Arita [Power and Industrial Systems R and D Div., Hitachi, Ltd., Hitachi, Ibaraki (Japan)

    1997-09-01

    A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators` demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators` demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs.

  9. Radiation effects in organic paints of a Boiling Water Reactor

    International Nuclear Information System (INIS)

    Jimenez Romo, J.A.; Olea Nader, M.A.

    1987-01-01

    The coatings on a BWR are used as a protection for the building and equipments from corrosion and contamination by radionuclides. The purpose of this work is to test this kind of coatings by simulating real absorbed doses in 40 years of use plus a nuclear accident (LOCA). Standards said that irradiation should be made with gamma radiation. In this work it's suggested to irradiate with electrons simulating secondary radiation produced on the interaction gamma-matter, and protons simulating the damage caused by the interaction neutron-matter. It's also suggested a new kind of adhesion test for coatings that gives a quantitative measure all other tests are qualitative. Two types of coatings were tested: Modified Phenolic and Epoxic both kinds had a very satisfactory performance in all the tests. The maximum dose accumulated by the coatings was 450 Mrad and the minimum 50 Mrad. The dose rates were: gamma in between 0.4 Mrad/hr and 1.0 Mrad/hr; protons and electrons between 500 Mrad/hr and 4000 Mrad/hr. Other important fact is that a calibration was made for a polymer to be used as a high dose dosimeter, these new dosimeters can measure doses between 10 Mrad and 100 Mrad not depending on the dose rate. (author)

  10. Development of alarm handling methods for boiling water reactors

    International Nuclear Information System (INIS)

    Ohga Yukiharu; Seki Hiroshi; Arita Setsuo

    1997-01-01

    A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators' demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators' demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs

  11. Construction of the advanced boiling water reactor in Japan

    International Nuclear Information System (INIS)

    Natsume, Nobuo; Noda, Hiroshi

    1996-01-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7

  12. State of development progress of advanced boiling water reactors

    International Nuclear Information System (INIS)

    Tomono, Katsuya

    1982-01-01

    Advanced BWRs being developed at present are those aiming at the improvement of reliability and safety, the reduction of radiation exposure, the improvement of operation performance and capacity ratio of plants, and the heightening of economical efficiency by concentrating the experience and excellent technology of BWR manufacturers in the world. Now in Japan, the independence with Japanese technology is possible in almost all fields of nuclear power generation, and the improvement and standardization project is in progress to obtain the steady results. However, in order to pursue the most desirable BWRs conceivable at present, five BWR manufacturers in the world organized the Advanced Engineering Team in July, 1978, and performed the feasibility study of advanced BWRs for more than one year. Tokyo Electric Power Co., Inc., evaluated the report on the results, and judged that it is desirable to advance into the next stage aiming at the practical use of advanced BWRs. For the purpose, the electric power common research on advanced BWRs has been in progress, and the A-BWR project is to be examined in the third improvement and standardization project of MITI. The main technical features such as the coolant recirculation system of internal pump type, reinforced concrete containment vessels, fine motion control rod drive, improved core and fuel and others are explained. (Kako, I.)

  13. The availability of German nuclear power plants in an international comparison

    International Nuclear Information System (INIS)

    Fehndrich, W.; Kutsch, W.

    1975-01-01

    The availability of German nuclear power stations is discussed in comparison with European and American light-water nuclear power stations. Furthermore, the availability or utilization is shown as a function of the year of operation and of the unit size (the latter only for the USA), and the trends observed during the last 4 years are made evident. The unscheduled shutdowns of light-water nuclear power stations during the past 4 years are analysed on the basis of publications of the IAEA in Vienna and the ABE-Committees of the German Atomforum with special emphasis on the affected systems. (orig.) [de

  14. Elemental Germans Klaus Fuchs, Rudolf Peierls and the making of British nuclear culture 1939-59

    CERN Document Server

    Laucht, Christoph

    2012-01-01

    Christoph Laucht offers the first investigation into the roles played by two German-born emigre atomic scientists, Klaus Fuchs and Rudolf Peierls, in the development of British nuclear culture, especially the practice of nuclear science and the political implications of the atomic scientists' work, from the start of the Second World War until 1959.

  15. Identification of neutron noise sources in a boiling water reactor

    International Nuclear Information System (INIS)

    Sides, W.H. Jr.; Mathis, M.V.; Smith, C.M.

    1977-01-01

    Measurements were made at units 2 and 3 of the Browns Ferry Nuclear Power Plant in order to characterize the noise signatures of the neutron and process signals and to determine the usefulness of such signatures for anomaly detection in BWR-4s. Previous measurements and theoretical analyses of BWR noise by others were concerned with the determination of steam velocity and void fraction (using the local component of neutron noise) and with the sources of global noise. The work described is under a five-part program to develop a complete and systematic analysis and representation of BWR neutron and process noise through complementary measurements and stochastic model developments. The parts are: (1) recording as many neutron detector and process noise signals as are available in a BWR-4; (2) reducing these data to noise signatures in order to perform an empirical analysis of these signatures, and documenting the relationships between the signals from spatially separated neutron detectors and between neutron and process variables; (3) developing spatially dependent neutronic models coupled with thermal-hydraulic models to aid in interpreting the observed relationships among the measured noise signatures, (4) comparing measured noise signatures with model predictions to obtain additional insight into BWR-4 dynamic behavior and to validate the models; and (5) using these models to predict the sensitivity of noise monitoring for detection, surveillance, and diagnosis of postulated in-core anomalies in BWRs. The paper describes the procedures used to obtain the noise recordings and presents initial empirical analysis and observations pertaining to the noise signatures and the relationships between several noise variables in the 0.01- to 1-Hz range. The mathematical models have not been developed sufficiently to report theoretical results or to compare measured spectra with model predictions at this time

  16. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    International Nuclear Information System (INIS)

    Boing, L.E.

    1989-12-01

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs

  17. Power distribution monitoring system in the boiling water cooled reactor core

    International Nuclear Information System (INIS)

    Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

    1987-01-01

    Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

  18. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.

    1989-12-01

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs.

  19. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    Kleiss, J.

    1983-01-01

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  20. Uranium from German nuclear power projects of the 1940s - a nuclear forensic investigation

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, Klaus; Wallenius, Maria; Luetzenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; Belle, Pieter van; Varga, Zsolt [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Buda, Razvan; Erdmann, Nicole [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Institut fuer Kernchemie, Universitaet Mainz (Germany); Kratz, Jens-Volker; Trautmann, Norbert [Institut fuer Kernchemie, Universitaet Mainz (Germany); Fifield, L. Keith; Tims, Stephen G. [Department of Nuclear Physics, Research School of Physics and Engineering, The Australian National University, Canberra, ACT (Australia); Froehlich, Michaela B. [Department of Nuclear Physics, Research School of Physics and Engineering, The Australian National University, Canberra, ACT (Australia); Universitaet Wien, Fakultaet fuer Chemie, Institut fuer Anorganische Chemie, Vienna (Austria); Steier, Peter [Universitaet Wien, Fakultaet fuer Physik, Isotopenforschung und Kernphysik, Vienna (Austria)

    2015-11-02

    Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel. Through measurement of the {sup 230}Th/{sup 234}U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the {sup 87}Sr/{sup 86}Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of {sup 236}U and {sup 239}Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  1. Uranium from German nuclear power projects of the 1940s - a nuclear forensic investigation

    International Nuclear Information System (INIS)

    Mayer, Klaus; Wallenius, Maria; Luetzenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; Belle, Pieter van; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L. Keith; Tims, Stephen G.; Froehlich, Michaela B.; Steier, Peter

    2015-01-01

    Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel. Through measurement of the 230 Th/ 234 U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the 87 Sr/ 86 Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of 236 U and 239 Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  2. The Brussels Nuclear Ship Convention and its impact on the German Atomic Energy Act

    International Nuclear Information System (INIS)

    Bernaerts, A.

    1976-01-01

    Although the Brussels Nuclear Ship Liability Convention of 1962, whose incorporation into German law was agreed upon by the German Federal Parliament (the Bundestag) in 1975, will have only minor international significance once it has entered into force, it will yet provide a new legal basis for the nuclear liability of the German vessel N.S. 'Otto Hahn' for the periods of her stay within the Federal Republic of Germany. However, there is no smooth concurrence of the Brussels Convention and the German Atomic Energy Act adapted to the Paris Liability Convention in 1975. This means that a number of questions still need to be resolved with respect to the protection of victims and the liability of shipowners. (orig.) [de

  3. Use of adaptive diffusion theory based monitors in optimizing boiling water reactor core designs

    International Nuclear Information System (INIS)

    Congdon, S.P.; Martin, C.L.; Crowther, R.L.

    1988-01-01

    Three-dimensional coarse mesh models are routinely used to predict the performance of boiling water reactors. In the adaptive monitory model, the three-dimensional solutions are permanently adapted to incore probe data. The corrections resulting from the adaptive process lead to reliable predictions of future reactor states. The corrections can also be carried forward to future operating cycles. This can shorten the time required to introduce an validate new design and operating strategy improvements. (orig.) [de

  4. Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

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

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  6. Evaluation of damages of airplane crash in European Advanced Boiling Water Reactor (EU-ABWR)

    International Nuclear Information System (INIS)

    Kamei, Kazuhiro; Tanoue, Tetsuharu; Kataoka, Kazuyoshi; Jimbo, Masakazu

    2011-01-01

    European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash (APC), severe accident mitigation systems, N+2 principle in safety systems and a large output of 1600 MWe. Thanks to above mentioned features, EU-ABWR's design objectives and principles are consistent with safety requirements in an European market. In this paper, evaluation of damages induced by APC has been summarized. (author)

  7. Determining the void fraction in draught sections of a boiling water cooled reactor

    International Nuclear Information System (INIS)

    Fedulin, V.N.; Barolomej, G.G.; Solodkij, V.A.; Shmelev, V.E.

    1987-01-01

    Consideration is being given to the problem of improving methods for calculation of the void fraction in large channels of cooling system of the boiling water cooled reactor during two-phase unsteady flow. Investigation of the structure of two-phase flow was conducted in draught section of the VK-50 reactor (diameter D=2 m, height H=3). The method for calculation of the void fraction in channels with H/D ratio close to 1 is suggested

  8. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    Fuentes C, P.

    2003-01-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O 2 ; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  9. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    Behrooz, A.

    2008-01-01

    The two-phase flow instability is of interest for the design and operation of many industrial systems such as boiling water reactors (BWRs), chemical reactors, and steam generators. In case of BWRs, the flow instabilities are coupled to the power instabilities via neutronic-thermal hydraulic feedbacks. Since these instabilities produce also local pressure oscillations, the coolant flashing plays a very important role at low pressure. Many frequency-domain codes have been used for two-phase flow stability analysis of thermal hydraulic industrial systems with particular emphasis to BWRs. Some were ignoring the effect of the local pressure, or the effect of 3D power oscillations, and many were not able to deal with the neutronics-thermal hydraulics problems considering the entire core and all its fuel assemblies. The new frequency domain tool uses the best available nuclear, thermal hydraulic, algebraic and control theory methods for simulating BWRs and analyzing their stability in either off-line or on-line fashion. The novel code takes all necessary information from plant files via an interface, solves and integrates, for all reactor fuel assemblies divided into a number of segments, the thermal-hydraulic non-homogenous non-equilibrium coupled linear differential equations, and solves the 3D, two-energy-group diffusion equations for the entire core (with spatial expansion of the neutron fluxes in Legendre polynomials).It is important to note that the neutronics equations written in terms of flux harmonics for a discretized system (nodal-modal equations) generate a set of large sparse matrices. The eigenvalue problem associated to the discretized core statics equations is solved by the implementation of the implicit restarted Arnoldi method (IRAM) with implicit shifted QR mechanism. The results of the steady state are then used for the calculation of the local transfer functions and system transfer matrices. The later are large-dense and complex matrices, (their size

  10. SWR 1000: the main design features of the advanced boiling water reactor with passive safety systems

    International Nuclear Information System (INIS)

    Carsten, Pasler

    2007-01-01

    The SWR-1000 (1000 MW) is a boiling water reactor whose economic efficiency in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies have been enlarged from a 10*10 rod array to a 12*12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving the control rods, as well as allowing spectral-shift and stretch-out operation. The plant safety concept is based on a combination of passive safety systems and a reduced number of active safety systems. All postulated accidents can be controlled using passive systems alone. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. The SWR-1000 is compliant with international nuclear codes and standards, and is also designed to withstand

  11. Plant life extensions for German nuclear power plants? Controversial discussion on potential electricity price effects

    International Nuclear Information System (INIS)

    Matthes, Felix C.; Hermann, Hauke

    2009-06-01

    The discussions on electricity price effects in case of the plant life extension of German nuclear power plants covers the following topics: (1) Introduction and methodology. (2) Electricity generation in nuclear power plants and electricity price based on an empirical view: electricity generation in nuclear power plants and final consumption price for households and industry in the European Union; electricity generation in nuclear power plants and electricity wholesale price in case of low availability of nuclear power plants in Germany; comparison of electricity wholesale prices in Germany and France. (3) Model considerations in relation to electricity prices and nuclear phase-out. (4) Concluding considerations.

  12. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    International Nuclear Information System (INIS)

    1997-05-01

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff's review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff's review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design

  13. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-05-01

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

  14. German risk study on nuclear power plants. Vol. 3

    International Nuclear Information System (INIS)

    Albertz, V.; Gueldner, W.; Hoemke, P.; Hoertner, H.; Lindauer, E.; Linden, J. von; Preischl, W.; Reichart, G.; Roehrs, W.

    1980-01-01

    This appendix discusses those studies which were carried out in phase A of the German Risk Study on the following problems: occurrence frequencies of initiating events, evaluation of the reactor pressure vessel, failure rates and failure probabilities for components, probabilities for human error. The appendix describes the general procedure used in determining these parameters. In addition, it gives the individual values used and the sources employed to determine these values. (orig.) [de

  15. SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

    International Nuclear Information System (INIS)

    Brettschuh, Werner

    2006-01-01

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

  16. SWR 1000: A Next-Generation Boiling Water Reactor Ready for Deployment

    International Nuclear Information System (INIS)

    Brettschuh, W.

    2006-01-01

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 design is particularly suitable for countries whose power systems do not include any large power plants. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (IandC) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10x10 rod array to a 12x12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free start-up, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving

  17. Predicted effect of power uprating on the water chemistry of commercial boiling water reactors

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Chu, Charles F.; Chang Ching

    2009-01-01

    The approach of power uprating has been adopted by operators of light water reactors in the past few decades in order to increase the power generation efficiency of nuclear reactors. The power uprate strategy is apparently applicable to the three nuclear reactors in Taiwan as well. When choosing among the three types of power uprating, measurement uncertainty, stretch power uprating, and extended power uprating, a deliberate and thorough evaluation is required before a final decision and an optimal selection can be made. One practical way of increasing the reactor power is to deliberately adjust the fuel loading pattern and the control rod pattern and thus to avoid replacing the primary coolant pump with a new one of larger capacity. The power density of the reactor will increase with increasing power, but the mass flow rate in the primary coolant circuit (PCC) of a light water reactor will slightly increase (usually by less than 5 %) or even remain unchanged. Accordingly, an uprated power would induce higher neutron and gamma photon dose rates in the reactor coolant but have a minor or no effect on the mass flow rate of the primary coolant. The radiolysis product concentrations and the electrochemical corrosion potential (ECP) values differ largely in the PCC of a boiling water reactor (BWR). It is very difficult to measure the water chemistry data directly at various locations of an actual reactor. Thus the impact of power uprating on the water chemistry of a BWR operating under hydrogen water chemistry (HWC) can only be theoretically evaluated through computer modelling. In this study, the DEMACE computer code was modified to investigate the impact of power uprating on the water chemistry under a fixed mass flow rate in the primary coolant circuit of a BWR/6 type plant. Simulations were carried out for hydrogen concentrations in feedwater ranging from 0.0 to 2.0 mg . kg -1 and for power levels ranging from 100 % to 120 %. The responses of water chemistry and ECP

  18. The opening of the debates on the extension of German nuclear plant lifetime

    International Nuclear Information System (INIS)

    Guidez, J.

    2010-01-01

    After having recalled that the German nuclear plants were to be progressively stopped by 2021, the author reports that the political coalition has thrown this Agenda back into question, because the development of alternative energies is insufficient to compensate the programmed closing down, and the German government wants to avoid the use of thermal plants which are generating CO 2 emissions. Negotiations are underway between the German government and the four electric companies in charge of different plants. It appears that even plants which were to be stopped in 2010, could keep on operating. The report contains tables indicating, for each nuclear plant, its type, its operator, its power, its dates of beginning and programmed end of commercial activity, its production variation in time. Some comments are given about already stopped reactors (Brunsbuttel and Krummel) which faced some problems in the past, and about waste storing sites (Asse, Gorleben, Morsleben, Konrad)

  19. The Nuclear Safety Convention - does it confirm existing German law, and update international law?

    International Nuclear Information System (INIS)

    Lindemann, C.

    1995-01-01

    Some selected examples are discussed that are intended to answer the question of whether the NSC in its essence represents a development in confirmation of existing German nuclear law, and whether, assuming its coming into effect, this Convention will mean a step forward in the development of international law. The author examines the value of this codification of international law as such, and some of the obligations and standards such as retrofitting measures or shutdown of reactors below safety standard, and continues with briefly discussing the relationship between the NSC and nuclear liability law, the planned provisions for radiological protection in Art. 15, and the obligations for transboundary notification of safety-relevant events. These stipulations are analysed in comparison to existing international law, and with a view to their implementation under German law. Some provisions of the NSC that are based on standards of international technical guidance are compared with German regulatory guides. (orig./HP) [de

  20. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    International Nuclear Information System (INIS)

    Nunez-Carrera, Alejandro; Francois, Juan Luis; Martin-del-Campo, Cecilia; Espinosa-Paredes, Gilberto

    2005-01-01

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the 233 U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235 U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly

  1. Variation of the effectiveness of hydrogen water chemistry in a boiling water reactor during power coastdown operations

    International Nuclear Information System (INIS)

    Yeh Tsungkuang; Wang Meiya; Chu, Charles F.; Chang Ching

    2009-01-01

    A theoretical model was adapted to evaluate the impact of power coastdown on the water chemistry of a commercial boiling water reactor (BWR) in this work. In principle, the power density of a nuclear reactor upon a power level decrease would immediately be lowered, followed by water chemistry variations due to reduced radiolysis of water and extended coolant residence times in the core and near-core regions. It is currently a common practice for a commercial BWR to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal feedwater hydrogen concentration may be different after a power coastdown is implemented in a BWR. A computer code DEMACE was used in the current study to investigate the impact of various power coastdown levels on major radiolytic species concentrations and electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of a domestic reactor operating under either normal water chemistry or HWC. Our analyses indicated that under a rated core flow rate the chemical species concentrations and the ECP did not vary monotonously with decreases in reactor power level at a fixed feedwater hydrogen concentration. In particular, ECP variations basically followed the patterns of hydrogen peroxide in the select regions and exhibited high values at power level of 90% for Reactor X. (author)

  2. Azcaxalli: A system based on Ant Colony Optimization algorithms, applied to fuel reloads design in a Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel-Estrada, Jaime, E-mail: jaime.esquivel@fi.uaemex.m [Facultad de Ingenieria, Universidad Autonoma del Estado de Mexico, Cerro de Coatepec S/N, Toluca de Lerdo, Estado de Mexico 50000 (Mexico); Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Ortiz-Servin, Juan Jose, E-mail: juanjose.ortiz@inin.gob.m [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Castillo, Jose Alejandro; Perusquia, Raul [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico)

    2011-01-15

    This paper presents some results of the implementation of several optimization algorithms based on ant colonies, applied to the fuel reload design in a Boiling Water Reactor. The system called Azcaxalli is constructed with the following algorithms: Ant Colony System, Ant System, Best-Worst Ant System and MAX-MIN Ant System. Azcaxalli starts with a random fuel reload. Ants move into reactor core channels according to the State Transition Rule in order to select two fuel assemblies into a 1/8 part of the reactor core and change positions between them. This rule takes into account pheromone trails and acquired knowledge. Acquired knowledge is obtained from load cycle values of fuel assemblies. Azcaxalli claim is to work in order to maximize the cycle length taking into account several safety parameters. Azcaxalli's objective function involves thermal limits at the end of the cycle, cold shutdown margin at the beginning of the cycle and the neutron effective multiplication factor for a given cycle exposure. Those parameters are calculated by CM-PRESTO code. Through the Haling Principle is possible to calculate the end of the cycle. This system was applied to an equilibrium cycle of 18 months of Laguna Verde Nuclear Power Plant in Mexico. The results show that the system obtains fuel reloads with higher cycle lengths than the original fuel reload. Azcaxalli results are compared with genetic algorithms, tabu search and neural networks results.

  3. Five years of partnership programme between German and East European nuclear power plant operators

    International Nuclear Information System (INIS)

    Kraemer, P.

    1995-01-01

    In a jointly adopted communique in August 1992 the regulations governing the partnerships were laid down. The following principles are given priority: bilateral exchange of operating experience on the basis of equal partnership; the knowledge gained from the exchange of experience is used to improve the operation and safety of the respective partner plant; the partnership programme is free from commercial interests and does not include transfer of funds. Since the establishment of the first partnership relations (Biblis) with an east european nuclear power plant (Balakowo) late in April 1990, a total of 126 reciprocal visits was paid, 82 of which to German nuclear power plants. The balance of the efforts made by German nuclear power plant operators to integrate east european nuclear plant operators can be considered positive. (orig./HP) [de

  4. On Stability of Natural-circulation-cooled Boiling Water Reactors during Start-up (Experimental Results)

    International Nuclear Information System (INIS)

    Manera, A.; Van der Hagen, T.H.J.J.

    2002-01-01

    The characteristics of flashing-induced instabilities, which are of importance during the start-up phase of natural-circulation Boiling Water Reactors (BWRs), are studied. Experiments at typical start-up conditions (low power and low pressure) are carried out on a steam/water natural circulation loop. The mechanism of flashing-induced instability is analyzed in detail and it is found that non-equilibrium between phases and enthalpy transport plays an important role in the instability process. Pressure and steam volume in the steam dome are found to have a stabilizing effect. The main characteristics of the instabilities have been analyzed. (authors)

  5. Investigation of boiling water reactor stability and limit-cycle amplitude

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.A.; Euler, J.A.

    1991-01-01

    Galerkin's method has been applied to a boiling water reactor (BWR) dynamics model consisting of the point kinetics equations, which describe the neutronics, and a feedback transfer function, which describes the thermal hydraulics. The result is a low-order approximate solution describing BWR behavior during small-amplitude limit-cycle oscillations. The approximate solution has been used to obtain a stability condition, show that the average reactor power must increase during limit-cycle oscillations, and qualitatively determine how changes in transfer function values affect the limit-cycle amplitude. 6 refs., 2 figs., 2 tabs

  6. Dynamic analysis of the condensate feedwater system in boiling water reactor plants

    International Nuclear Information System (INIS)

    Tanji, J.; Omori, T.

    1982-01-01

    The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

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

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  8. An Analysis of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Persson, P.

    1963-06-01

    A method of predicting the burnout conditions for flow of boiling water in vertical round ducts is presented. The analysis predicts that the burnout conditions are independent of the L/d-ratio and the inlet temperature, and that the burnout steam quality decreases with increasing surface heat flux and increasing mass velocity. It was also found that the burnout steam quality at low pressures increases with the pressure and reaches a maximum at approximately 70 kg/cm, and thereafter decreases with a further increase of the pressure. The theoretical result compares very well with experimental data from different sources

  9. An Analysis of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Persson, P

    1963-06-15

    A method of predicting the burnout conditions for flow of boiling water in vertical round ducts is presented. The analysis predicts that the burnout conditions are independent of the L/d-ratio and the inlet temperature, and that the burnout steam quality decreases with increasing surface heat flux and increasing mass velocity. It was also found that the burnout steam quality at low pressures increases with the pressure and reaches a maximum at approximately 70 kg/cm, and thereafter decreases with a further increase of the pressure. The theoretical result compares very well with experimental data from different sources.

  10. Decontamination flange film characterization for a boiling water reactor under hydrogen water chemistry

    International Nuclear Information System (INIS)

    Baston, V.F.; Garbauskas, M.F.; Bozeman, J.

    1996-01-01

    Stainless steel artifacts removed from a boiling water reactor class 4 plant that operated under hydrogen water chemistry and experienced a difficult decontamination were submitted for oxide film characterization. The results reported for the corrosion film composition and structure are consistent with existing theoretical concepts for stainless steel corrosion, spinel structure site preferences (octahedral or tetrahedral) for transition metal ions, and potential-pH diagrams. The observed zinc effects on film stability and lower cobalt incorporation are also consistent with these theoretical concepts

  11. Instrumenting a pressure suppression experiment for a Mark I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1978-01-01

    A 1 / 5 -scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of the dynamic loading of the pressure-suppression system

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

    International Nuclear Information System (INIS)

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

    1993-08-01

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

  13. Overview of activities for the reduction of dose rates in Swiss boiling water reactors

    International Nuclear Information System (INIS)

    Alder, H.P.; Schenker, E.

    1993-01-01

    Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

  14. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

  15. A potential of boiling water power reactors with a natural circulation of a coolant

    International Nuclear Information System (INIS)

    Osmachkin, V.S.; Sokolov, I.N.

    1998-01-01

    The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

  16. Imagined Modernity. Nuclear Power and West German Society in the 1960s

    International Nuclear Information System (INIS)

    Metzler, G.

    2012-01-01

    In 1969 the first commercial nuclear power plant went into operation in West Germany, making the country a latecomer (compared to other Western countries) in the civilian use of nuclear energy. Yet, in the late 1950s and 1960s, nuclear energy played a major part in public debates on energy supply, on science and technology policy, and the relationship between state and the economy. The civilian use of atomic energy also served as a source of national identity and a projection of modernity. The paper seeks to connect analyses of technical, economic, political, and cultural developments, in order to assess the overall impact of nuclear power on West German society. (author)

  17. The trend and precept of present nuclear energy policy in German

    International Nuclear Information System (INIS)

    Murakami, Takashi

    2000-01-01

    In German, the coalition political power of SPD and Gruene decided the present nuclear policy on 20th October 1998. It consisted of three steps: the first step is amendment of the present Atomic Energy Law, the second step the consensus conference with the power industry for one year and the third step is enactment of law to decrease nuclear power with compensation. The first consensus conference opened 26th January 1999. The content was that the stop of reprocessing did not decide the limit and the atomic power plants were operated until obtaining consensus. Then, Muller Secretary of Commerce and the power industry carried out unofficial meeting and decided the following: limit of operating atomic power plants is 35 years (until 2024), reprocessing can be done until 2004 and the undertakers do not request compensation to the governmental agencies. However the environmental secretary and Gruene objected to them. 570 of German scientists and technologists published memorandum for discussing the present conditions of German and the earth. England and France opposite limit of reprocessing spent fuel. A joint concern between Siemens (German) and Framatome (France) is going to establish. The point under discussion about present power policy is selection of energy. There are many problems in German. (S.Y.)

  18. Lessons from American-German nuclear power plant construction. Quality, safety and costs of an attempt to integrate American and German nuclear power plant technology

    International Nuclear Information System (INIS)

    Buchwald, K.

    1979-05-01

    The 1300 MW nuclear power plant at Muelheim-Kaerlich has been under construction since the beginning of 1975. It is being equipped with a pressurised water reactor which has been adapted to the German client's requirements and German licensing practice, based on a license held by Babcock and Wilcox USA (B and W). The problems which have arisen in making this adaptation are the result of different requirements in the USA and the Federal Republic of Germany which make it very difficult to integrate the two technologies. Full integration will almost certainly be impossible, but integration to the widest possible extent is important because it might mean both greater safety and reduced costs. In this article it is intended to show where the problems of integration lie and how they might perhaps be overcome. (author)

  19. Nuclear energy policy in the German Democratic Republic

    International Nuclear Information System (INIS)

    Kahlert, J.

    1988-01-01

    The book investigates the question whether the peaceful use of nuclear energy is a technology accepted and mastered by society in the G.D.R., i.e. in a system of real socialism. The author first reviews the history of nuclear technology and its changing role from a source of hope and trust in progress to a fill-in. Then the question is discussed whether the combination of nuclear technology and socialism has proven the world to be a successful symbiosis. It is shown that discussions about the advantages or disadvantages of nuclear technology, actually a means of forming public opinion, are kept behind closed doors and are by no means intended to include participation of the public. The author concludes that nuclear technology is not controlled by society. Its development was determined by feasibility speculations, economic pretentions, and vague prognostics rather than by sound projecting and planning on the basis of technical, economic and energy policy principles. Real socialism, too, puts an unmeasurable risk on society by supporting nuclear technology. Public opinion making by no means allows controversial discussions about assessment of nuclear technology, its value, or the acceptable risk associated with it. (orig./HSCH) [de

  20. Integrated plant safety assessment. Systematic Evaluation Program. La Crosse Boiling Water Reactor. Dairyland Power Cooperative, Docket No. 50-409. Final report

    International Nuclear Information System (INIS)

    1983-06-01

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review

  1. Integrated plant safety assessment: Systematic Evaluation Program. LaCrosse Boiling Water Reactor, Dairyland Power Cooperative, Docket No. 50-409

    International Nuclear Information System (INIS)

    1983-04-01

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addresed. Equipment and procedural changes have been identified as a result of the review

  2. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    International Nuclear Information System (INIS)

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path

  3. German Federal spendings on nuclear energy in 1989

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The portfolio of the BMFT (Federal Ministry of Research and Technology) covers under the competence of the Federal Government all activities in the field of nuclear science and engineering for peaceful uses of nuclear energy, reactor safety research, and research on non-nuclear energy sources and technology. The draft budget for 1989 shows a total expenditure of DM 7.65 billions in the section 30, portfolio of the BMFT. This is about 1.2% more than in the draft budget of 1988. Broken down into programmes, DM 1.853 billions are earmarked for energy research and technology (1988: DM 1.854 billions), of these DM 398.5 millions for the promotion of non-nuclear energy research and technology. (orig./UA) [de

  4. Analytic advanced development in the German Risk Study on Nuclear Power Plants

    International Nuclear Information System (INIS)

    1984-01-01

    The study carried out by the Ecological Institute (Oeko-Institut) deals with the difficulties encountered with the decision-making models and statistical models and methods of the German Risk Study on Nuclear Power Plants for the determination of care meltdown frequency and for the estimation of the behaviour of components in the event of a failure. Related deficiencies of the German Risk Study and the absence of completeness and lading representiveness of the course and causes of incidents considered are pointed out. The accident consequences model and the climatological-meteorological data base are considered non-representative. (RF) [de

  5. Operating experience of natural circulation core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    Kullberg, C.; Jones, K.; Heath, C.

    1993-01-01

    General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

  6. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  7. The German-German history of the nuclear power plant Greifswald. Nuclear power between east and west. 2. ed.; Die deutsch-deutsche Geschichte des Kernkraftwerkes Greifswald. Atomenergie zwischen Ost und West

    Energy Technology Data Exchange (ETDEWEB)

    Hoegselius, Per [Technische Hochschule Stockholm (Sweden). Bereich Gesellschaft, Wissenschaft und Technik

    2015-07-01

    The historical study covers the chapters The nuclear power plant Greifswald; Lubmin shortly before the ''Wende'' 1989; the German ''Wende''; from the last vote for the ''Volkskammer (parliament of the German Democratic Republic) to the German reunification; Lubmin in reunified Germany; conclusions and perspectives. In the attachment technical data about the reactors WWER-440/W-230 are summarized, including a list of WWERs in the former eastern bloc countries.

  8. The future of nuclear power: An IAEA perspective. Address at the German atomic forum, Berlin, 27 January 2000

    International Nuclear Information System (INIS)

    ElBaradei, M.

    2000-01-01

    In his address at the German Atomic Forum (Berlin, 27 January 2000), the Director General of the IAEA presented a survey of the prospects for nuclear power as seen from the perspective of the IAEA. The speech focussed on the following aspects: nuclear power and world energy requirements, nuclear and radiation safety, competitiveness of nuclear power, and public support

  9. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    Directory of Open Access Journals (Sweden)

    A. Schwenk-Ferrero

    2013-01-01

    Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

  10. German Democratic Republic State system of accounting for and control of nuclear material

    International Nuclear Information System (INIS)

    Roehnsch, W.; Gegusch, M.

    1976-01-01

    The system of accountancy for and control of nuclear material in the German Democratic Republic (GDR) with its legal bases and components is embedded in the overall State system of protection in the peaceful uses of nuclear energy. As the competent State authority, the Nuclear Safety and Radiation Protection Board of the GDR is also responsible for meeting the GDR's national and international tasks in the control of nuclear material. At enterprise level, the observance of all safety regulations for nuclear material, including the regulations for the control, is within the responsibility of managers of establishments, which are in any way concerned with the handling of nuclear material. To support managers and to function as internal control authorities, nuclear material officers have been appointed in these establishments. Design information, operating data, physical inventory of nuclear material and the respective enterprise records and reports are subject to State control by the Nuclear Material Inspectorate of the Nuclear Safety and Radiation Protection Board. This Inspectorate keeps the central records on nuclear material, forwards reports and information to, and maintains the necessary contacts with, the IAEA. For the nuclear material in the GDR four material balance areas have been established for control purposes. To rationalize central recording and reporting, electronic data processing is increasingly made use of. In a year-long national and international control of nuclear material, the State control system has stood the test and successfully co-operates with the IAEA. (author)

  11. German risk study for nuclear power plants. Vol. 5

    International Nuclear Information System (INIS)

    Bracht, K.F.; Javeri, V.; Keusenhoff, J.; Meier, S.; Roehrs, W.; Scharfe, A.; Tiltmann, M.; Mayinger, F.

    1980-01-01

    This appendix contains the analyses on core melt accidents carried out for the German risk study. These accidents constitute a major portion of the event sequences relevant to the risk. It deals, in particular, with the processes involved in the reactor core melt, the behaviour of the containment vessel including the possible types of failures, and the consequences of possible steam explosions. The investigations were performed for limiting cases for which a complete failure of the emergency core cooling and residual heat removal system is anticipated. Under these conditions, the reactor core melts down and penetrates the reactor pressure vessel bottom. In the course of the following interaction between core melt and concrete foundation, the melt is assumed to come into contact with the sump water. The evaporation of the sump water leads to an overpressure failure of the containment vessel within approximately one day after accident initiation. A destruction of the containment as a consequence of a steam explosion is considered highly unprobable. (orig.) [de

  12. Nuclear safety and radiation protection in the German Democratic Republic

    International Nuclear Information System (INIS)

    Sitzlack, G.; Scheel, H.

    1976-01-01

    The radiation protection organization in the GDR is outlined laying emphasis on the tasks of the National Board of Nuclear Safety and Radiation Protection. In addition to the basic tasks, the various forms of radiation protection monitoring, the management of radioactive wastes, and international responsibilities are briefly explained. (author)

  13. Reliability data of fire protection equipment and features in German nuclear power plants

    International Nuclear Information System (INIS)

    Roewekamp, M.; Riekert, T.; Sehrbrock, W.

    1997-01-01

    In order to perform probabilistic fire safety analyses, a comprehensive data base is needed including physical characteristics of fire compartments and their inventory, fire occurrence frequencies, technical reliability data for all fire-related equipment, human actions and human error probabilities, etc. In order to provide updated and realistic reliability data, the operational behaviour of different fire protection features in two German nuclear power plants was analysed in the framework of the study presented here. The analyses are based on the examination of reported results of the regular inspection and maintenance programs for nuclear power plants. Besides a plant specific assessment of the reliability data a generic assessment for an application as input data for fault tree analyses in the framework of probabilistic risk studies for other German plants was carried out. The analyses of failures and unavailabilities gave the impression that most of them are single failures without relevance for the plant safety. The data gained from NPPs were compared to reliability data of the German insurance companies for the same protection features installed in non-nuclear installations and to older nuclear specific reliability data. This comparison showed up a higher reliability. (orig.) [de

  14. Environmentalism and nuclear power: Anglo-German comparison

    International Nuclear Information System (INIS)

    Boehmer-Christiansen, Sonja

    1992-01-01

    The paper deals with impacts of emission reduction targets on fuel policy for electricity, and vice versa. It does so by comparing policy responses in two different institutional contexts. Supportive links were created, in Germany, between the reduction of acid emissions from power stations and the protection of nuclear power. This higher cost option was justified to an 'aware' public on environmental grounds. Politically this required a tacit alliance between coal-based and nuclear industries realising that the well-being of both depended on public support, the common challenge being fossil fuel suppliers in world markets. Cooperation with the State was therefore desirable and, in Germany, available. The British story is very different and not without irony. Similarities between the energy politics of acid rain and the global warming/climate change debate are explored. It is concluded that institutions and linkages between them matter very much. (author)

  15. NS OTTO HAHN - the first German nuclear ship

    International Nuclear Information System (INIS)

    1981-01-01

    The NS OTTO HAHN is the first and only European nuclear propelled cargo and research vessel. She entered service in 1968 and was operated for 11 years without any technical failure. The essential experience and know-how about the nuclear propulsion unit is available now. Therefore the ship was decommissioned in 1979. Until the end of 1981 all activated and contaminated components will be removed and decontaminated. The ship can then be released for conventional utilization. In this book the NS OTTO HAHN is described in detail and the experiences of operation as well as research and development activities are reported. All earlier publications of GKSS on the same subject are listed. (orig.) [de

  16. Organization of radiation protection in German nuclear power stations

    International Nuclear Information System (INIS)

    1989-01-01

    Using the operating handbooks of the nuclear power stations in West Germany, an examination was carried out of how far the existing organisational structure for radiation protection fulfils the requirements for protection and whether a standardisation of the organisation would provide improvements for the protection of the personnel and for the practicability of the radiation protection organisation. In particular, the parts 'Personnel operating organisation', 'Radiation protection order' and 'Maintenance order' of the operating handbook were evaluated and an audit was made of the radiation protection organisation. In general, the result of the assessment is that the organisation of radiation protection does not contradict the orders, guidelines and regulations in any of the nuclear power stations examined. Corresponding to the possibilities of regulating details of the radiation protection organisation within the undertaking, the target of 'protection of the personnel against radioactive irradiation' is achieved by the various organisation structures which are largely equal to the given example. (orig./HP) [de

  17. Nuclides from Chinese nuclear weapons tests in German kale

    International Nuclear Information System (INIS)

    Schelenz, R.; Fischer, E.

    1979-01-01

    Fission products (Be-7, Ce-141, Ce-137, J-131, Nd-147, Ru-103, Zr-95) of two atmospheric nuclear explosions in the People's Republic of China (September 17, 1977 and March 15, 1978) were detected in kale in the Federal Republic of Germany 12 days after the detonations. The slight increase of radioactivity was negligible with respect to the radiation exposure of man even if large amounts of vegetables were consumed. (orig.) [de

  18. German foreign trade in nuclear products 1978-1980

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    This statistics of the imports and exports of nuclear products of the Federal Republic of Germany in 1980, compiled by the Federal Minister for Research and Technology (with reference figures for 1978 and 1979) shows that imports have decreased by 23,2% (DM 500,000 million) to DM 1,500,000 million. In contrast between 1978 and 1979 there was a 26,2% increase (DM 400,000 million). (orig./UA) [de

  19. Nuclides from Chinese nuclear weapons in German kale

    International Nuclear Information System (INIS)

    Schelenz, R.; Fischer, E.

    1979-01-01

    Fission products (Be-7, Ce-141, Cs-137, J-131, Nd-147, Ru-103, Zr-95) of two atmospheric nuclear explosions in the People's Republic of China (September 17, 1977 and March 15, 1978) were detected in kale in the Federal Republic of Germany 12 days after the detonations. The slight increase of radio activity was negligible with respect to the radiation exposure of man even if large amounts of vegetables were consumed. (RW) [de

  20. Nuclides from Chinese nuclear weapons tests in German kale

    International Nuclear Information System (INIS)

    Schelenz, R.; Fischer, E.

    1979-01-01

    Fission products of two atmospheric nuclear explosions in the People's Republic of China (September 17, 1977 and March 15, 1978) were detected in kale in the Federal Republick of Germany 12 days after the detonations. The slight increase of radio activity was negligible with respect to the radiation exposure of man even if large amounts of vegetables were consumed. (orig.) 891 AJ/orig. 892 MB [de

  1. Notifiable events in German nuclear power plants in 1992

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    In 1992, there were 223 notifiable nuclear power plant incidents in the Federal Republic of Germany. 221 incidents belong to the category N (normal notification, INES 0), 2 incidents belong to the category E (immediate notification, INES 2). The report comprehensively lists all the incidents. No releases of radioactive materials above the, lincenced levels occupied and there were no hazards to persons and to the environment. (HP) [de

  2. Atmospheric discharges from nuclear facilities during decommissioning: German experiences

    Energy Technology Data Exchange (ETDEWEB)

    Braun, H.; Goertz, R.; Weil, L.

    1997-08-01

    In Germany, a substantial amount of experience is available with planning, licensing and realization of decommissioning projects. In total, a number of 18 nuclear power plants including prototype facilities as well as 6 research reactors and 3 fuel cycle facilities have been shut down finally and are at different stages of decommissioning. Only recently the final {open_quotes}green field{close_quotes} stage of the Niederaichbach Nuclear Power Plant total dismantlement project has been achieved. From the regulatory point of view, a survey of the decommissioning experience in Germany is presented highlighting the aspects of production and retention of airborne radioactivity. Nuclear air cleaning technology, discharge limits prescribed in licences and actual discharges are presented. As compared to operation, the composition of the discharged radioactivity is different as well as the off-gas discharge rate. In practically all cases, there is no significant amount of short-lived radionuclides. The discussion further includes lessons learned, for example inadvertent discharges of radionuclides expected not to be in the plants inventory. It is demonstrated that, as for operation of nuclear power plants, the limits prescribed in the Ordinance on Radiological Protection can be met using existing air cleaning technology, Optimization of protection results in public exposures substantially below the limits. In the frame of the regulatory investigation programme a study has been conducted to assess the airborne radioactivity created during certain decommissioning activities like decontamination, segmentation and handling of contaminated or activated parts. The essential results of this study are presented, which are supposed to support planning for decommissioning, for LWRs, Co-60 and Cs-137 are expected to be the dominant radionuclides in airborne discharges. 18 refs., 2 figs., 1 tab.

  3. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    Science.gov (United States)

    Gorman, Phillip Michael

    The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

  4. German offsite accident consequence model for nuclear facilities: further development and application

    International Nuclear Information System (INIS)

    Bayer, A.

    1985-01-01

    The German Offsite Accident Consequence Model - first applied in the German Risk Study for nuclear power plants with light water reactors - has been further developed with the improvement of several important submodels in the areas of atmospheric dispersion, shielding effects of houses, and the foodchains. To aid interpretation, the presentation of results has been extended with special emphasis on the presentation of the loss of life expectancy. The accident consequence model has been further developed for application to risk assessments for other nuclear facilities, e.g., the liquid metal fast breeder reactor (SNR-300) and the high temperature gas cooled reactor. Moreover the model have been further developed in the area of optimal countermeasure strategies (sheltering, evacuation, etc.) in the case of the Central European conditions. Preliminary considerations has been performed in connection with safety goals on the basis of doses

  5. Third Finnish-German seminar on nuclear waste management 1986

    International Nuclear Information System (INIS)

    Lamberg, L.

    1988-01-01

    The scope of the seminar was to provide an interdisciplinary forum for exchange of information and experiences in the field of nuclear waste management. The highlights of the seminar focused on the following topics: overall reviews, waste products, nearfield phenomena, site investigations, performance assessment and decommissioning. All together 20 papers were presented. Reviews, status reports and experimental studies dealt with general research programs and current research and development activities including regulatory aspects. Extensive discussions provided and opportunity to identify issues and options for further research

  6. 1997: Nuclear power in the German power industry

    International Nuclear Information System (INIS)

    Wnuk, A.

    1998-01-01

    Also in 1997, nuclear power contributed the largest share among all sources of energy to the supply of electricity in Germany, generating 170,3 TWh. This marks a considerable increase by almost 5% over the previous year's level. Nuclear power held a 31.0% share in the supply of electricity in Germany, and a 34.7% share in public supply. Hard coal, producing 143.1 TWh, contributed a share of 26.0%. Electricity generation from lignite decreased by 1.8%, reaching 141.7 TWh; its share is now 25.8%. Hydrocarbons hold a share in gross electricity generation of nearly 12% (to which oil merely contributes 1%), which is a low level by international standards. Hydroelectricity generation decreased by 3.6% down to 3.8%. The upward trend of the economy raised the electricity requirement in industry by 2.2%. Net consumption by private households decreased by 2.5%. For the year under review, the foreign trade balance in electricity showed an excess of exports over imports of 2.4 TWh. (orig.) [de

  7. 1996: nuclear power in the German power industry

    International Nuclear Information System (INIS)

    Wnuk, A.

    1997-01-01

    Also in 1996, nuclear power contributed the largest share among all sources of energy to the supply of electricity in Germany, generating 161,1 TWh. This marks a considerable increase by almost 5% over the previous year's level. Nuclear power held a 29.4% share in the supply of electricity in Germany, and a 33% share in public supply. Hard coal, producing 152.7 TWh, contributed a share of 27.7%. Electricity generation from lignite rose slightly by 1.2%, reaching 144.3 TWh; its share is now 26.2%. Hydrocarbons hold an 11.5% share in gross electricity generation (to which oil merely contributes 1.3%), which is a low level by international standards. After a continuous rise over the past few years, hydroelectricity generation for the first time showed a decrease by 10.5% because of adverse water conditions. The upward trend of the economy raised the electricity requirement in industry by 1.5%. As a consequence of the long spell of cold weather, net consumption by private households rose by 5.5%. For the year under review, the foreign trade balance in electricity for the first time after 1992 again showed an excess of exports over imports of 5.3 TWh. (orig.) [de

  8. Rise and fall of the German nuclear industry; Aufstieg und Fall der deutschen Atomwirtschaft

    Energy Technology Data Exchange (ETDEWEB)

    Radkau, Joachim [Bielefeld Univ. (Germany). Neuere Geschichte; Hahn, Lothar

    2013-02-01

    The book on the rise and fall of the German nuclear industry includes five chapters: (1) From the atomic project of the second world war to the ''peaceful atom''. (2) The ''peaceful atom'' as vision: the phase of speculations. (3) Achieved facts: the unplanned triumph of the light water reactor. (4) The internally suppressed risk excites the public. (5) From the creeping to the open fall.

  9. Electrochemical potential measurements in boiling water reactors; relation to water chemistry and stress corrosion

    International Nuclear Information System (INIS)

    Indig, M.E.; Cowan, R.L.

    1981-01-01

    Electrochemical potential measurements were performed in operating boiling water reactors to determine the range of corrosion potentials that exist from cold standby to full power operation and the relationship of these measurements to reactor water chemistry. Once the corrosion potentials were known, experiments were performed in the laboratory under electrochemical control to determine potentials and equivalent dissolved oxygen concentrations where intergranular stress corrosion cracking (IGSCC) would and would not occur on welded Type-304 stainless steel. At 274 0 C, cracking occurred at potentials that were equivalent to dissolved oxygen concentration > 40 to 50 ppb. With decreasing temperature, IGSCC became more difficult and only severely sensitized stainless steel would crack. Recent in-reactor experiments combined with the previous laboratory data, have shown that injection of small concentrations of hydrogen during reactor operation can cause a significant decrease in corrosion potential which should cause immunity to IGSCC. (author)

  10. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  11. Failure analysis of cracked head spray piping from the Dresden Unit 2 Boiling Water Reactor

    International Nuclear Information System (INIS)

    Diercks, D.R.; Dragel, G.M.

    1983-07-01

    Several sections of Type 304 stainless steel head spray piping, 6.25 cm (2.5 in.) in diameter, from the Dresden Unit 2 Boiling Water Reactor were examined to determine the nature and causes of coolant leakages detected during hydrostatic tests. Extensive pitting was observed on the outside surface of the piping, and three cracks, all located at a helical stripe apparently rubbed onto the outer surface of the piping, were also noted. Metallographic examination revealed that the cracking had initiated at the outer surface of the pipe, and showed it to be transgranular and highly branched, characteristic of chloride stress corrosion cracking. The surface pitting also appeared to have been caused by chlorides. A scanning electron microprobe x-ray analysis of the corrosion product in the cracks confirmed the presence of chlorides and also indicated the presence of calcium

  12. Corrosion products, activity transport and deposition in boiling water reactor recirculation systems

    International Nuclear Information System (INIS)

    Alder, H.P.; Buckley, D.; Grauer, R.; Wiedemann, K.H.

    1992-01-01

    The deposition of activated corrosion products in the recirculation loops of Boiling Water Reactors produces increased radiation levels which lead to a corresponding increase in personnel radiation dose during shut down and maintenance. The major part of this dose rate is due to cobalt-60. Based on a comprehensive literature study concerning this theme, it has been attempted to identify the individual stages of the activity build-up and to classify their importance. The following areas are discussed in detail: The origins of the corrosion products and of cobalt-59 in the reactor feedwaters; the consolidation of the cobalt in the fuel pins deposits (activation); the release and transport of cobalt-60; the build-up of cobalt-60 in the corrosion products in the recirculation loops. Existing models of the build-up of circuit radioactivity are discussed and the operating experiences from selected reactors are summarized. 90 refs, figs and tabs

  13. Approximation model of three-dimensional power distribution in boiling water reactor using neural networks

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2001-01-01

    Fast and accurate prediction of three-dimensional (3D) power distribution is essential in a boiling water reactor (BWR). The prediction method of 3D power distribution in BWR is developed using the neural network. Application of the neural network starts with selecting the learning algorithm. In the proposed method, we use the learning algorithms based on a class of Quasi-Newton optimization techniques called Self-Scaling Variable Metric (SSVM) methods. Prediction studies were done for a core of actual BWR plant with octant symmetry. Compared to classical Quasi-Newton methods, it is shown that the SSVM method reduces the number of iterations in the learning mode. The results of prediction demonstrate that the neural network can predict 3D power distribution of BWR reasonably well. The proposed method will be very useful for BWR loading pattern optimization problems where 3D power distribution for a huge number of loading patterns (LPs) must be performed. (author)

  14. TRAC-BD1: transient reactor analysis code for boiling-water systems

    International Nuclear Information System (INIS)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

  15. Effect of water impurities on stress corrosion cracking in a boiling water reactor

    International Nuclear Information System (INIS)

    Ljungbery, L.G.; Cubicciotti, D

    1985-01-01

    A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

  16. Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

    Science.gov (United States)

    Gordon, Barry; Garcia, Susan

    Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

  17. Investigation of the resonant power oscillation in the Halden Boiling Water Reactor by autoregressive modeling

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1980-01-01

    In the HBWR (Halden Boiling Water Reactor), there exists a resonant power oscillation with period about 0.04 Hz at power levels higher than about 9.5 MWt. While the resonant oscillation in not so large as to affect the normal reactor operation, it is significant, from the viewpoint of reactor diagnosis, to grasp its characteristics and find the cause. Noise analysis based on the autoregressive (AR) modeling technique has been made to reveal the driving source for this oscillation which led to the suggestion that it is attributed to the dynamic interference of heat exchange process between two parallel-connected steam transformers against the reactor. The present study demonstrates that the method used here is highly effective for tracing back to a noise source inducing the variation of quantities in a system, and also applicable to problems of reactor noise analysis and diagnosis. (author)

  18. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  19. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    International Nuclear Information System (INIS)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-01-01

    An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

  20. Passive containment cooling system performance in the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Gamble, R.E.; Yadigaroglu, G.

    1997-01-01

    The Simplified Boiling Water Reactor (SBWR) incorporates a passive system for decay heat removal from the containment in the event of a postulated Loss-of-Coolant Accident (LOCA). Decay heat is removed by condensation of the steam discharged from the reactor pressure vessel (RPV) in three condensers which comprise the Passive Containment Cooling System (PCCS). These condensers are designed to carry the heat load while transporting a mixture of steam and noncondensible gas (primarily nitrogen) from the drywell to the suppression chamber. This paper describes the expected LOCA response of the SBWR with respect to the PCCS performance, based on analysis and test results. The results confirm that the PCCS has excess capacity for decay heat removal and that overall system performance is very robust. 12 refs., 8 figs

  1. Review of boiling water reactor small break loss of coolant accidents

    International Nuclear Information System (INIS)

    Gururaj, P.M.; Dua, S.S.; Rao, A.S.

    1981-01-01

    This paper presents a review of the analytical and the experimental work performed by the General Electric Company to determine the performance of boiling water reactors (BWR) following postulated small break accidents (SBA). This review paper addresses the following issues: (1) the response of the BWR following small loss of inventory events; (2) methods of analysis and their justification; (3) necessity, if any, of operator action and the length of time available in which such action can be performed; and (4) operator interface following the SBA event. The results from these SBA studies for different BWR product lines show that even with the multiple system failures assumed, the BWR can successfully withstand an SBA. For a typical BWR/6, it takes the failure of 13 water delivery pumps to cause any significant core heatup. The only operator actions determined to be necessary are simple ones and ample time is available to the operator to perform these actions, if needed

  2. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Haden, T.H.J.J. van der; Zboray, R.; Manera, A.; Mudde, R.F.

    2002-01-01

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

  3. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    International Nuclear Information System (INIS)

    Schlereth, J.R.; Pennington, D.

    1996-01-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it's Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components

  4. Experimental determination of residual stress by neutron diffraction in a boiling water reactor core shroud

    International Nuclear Information System (INIS)

    Payzant, A.; Spooner, S.; Zhu, Xiaojing; Hubbard, C.R.

    1996-01-01

    Residual strains in a 51 mm (2-inch) thick 304L stainless steel plate have been measured by neutron diffraction and interpreted in terms of residual stress. The plate, measuring (300 mm) in area, was removed from a 6m (20-ft.) diameter unirradiated boiling water reactor core shroud, and included a multiple-pass horizontal weld which joined two of the cylindrical shells which comprise the core shroud. Residual stress mapping was undertaken in the heat affected zone, concentrating on the outside half of the plate thickness. Variations in residual stresses with location appeared consistent with trends expected from finite element calculations, considering that a large fraction of the residual hoop stress was released upon removal of the plate from the core shroud cylinder

  5. Modelling of crack chemistry in sensitized stainless steel in boiling water reactor environments

    International Nuclear Information System (INIS)

    Turnbull, A.

    1997-01-01

    An advanced model has been used to predict the chemistry and potential in a stress corrosion crack in sensitized stainless steel in a boiling water reactor (BWR) environment. The model assumes trapezoidal crack geometry, incorporates anodic reaction and cathodic reduction within the crack, and takes into account the limited solubility of cations in high temperature water. The results indicate that the crack tip potential is not independent of the external potential, and that the reactions on the walls of the crack must be included for reliable prediction. Accordingly, both the modelling assumptions of Ford and Andresen and of Macdonald and Urquidi-Macdonald, whilst having merit, are not fully satisfactory. Extended application of the model for improved prediction of stress corrosion crack growth rate is constrained by limitations in electrochemical data which are currently inadequate. (author)

  6. Stainless steels in boiling water reactors. Corrosion problems and possible solutions

    International Nuclear Information System (INIS)

    Combrade, P.; Desestret, A.; Leroy, F.; Coriou, H.

    1977-01-01

    In boiling water reactors, the heat-carrying water may have an up to 0.1 or even 0.2 ppm oxygen content, which can make it highly agressive at operating temperature for stainless steels subject to high physical stresses. Several metallurgical solutions can be considered, and in particular the use of stainless steels having a mixed austenitic-ferritic structure or of standard austenitic steels (18.10 or 18.10 Mo, such as AISI 304 and 316) with carefully controlled carbon and alloy element contents. The behavior of these steels during prolonged tests in water at 288 0 C with a 30 and even 100 ppm oxygen content turned out to be quite satisfactory [fr

  7. A method of simulating voids in experimental studies of boiling water reactors

    International Nuclear Information System (INIS)

    Down, H.J.; Dickie, J.; Fox, W.N.

    1963-11-01

    The coolant density in boiling water reactors may vary from 3 at pressures up to 1000 p.s.i. In order to study the effect of reduced water density on reactivity in unpressurized experimental systems, the effective water density is reduced by packing small beads of highly expanded polystyrene into the fuel clusters and flooding the interstices with water. Coolant densities of from 0.4 to 0.6 gm/cm 3 may be produced with the introduction of only about 0.4 gm/cm 3 of non-hydrogeneous material. This memorandum describes the production, properties and handling of polystyrene beads and the tests carried out to establish the validity of the technique. (author)

  8. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts (Part 2)

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Persson, P.; Nilsson, L.; Eriksson, O.

    1963-06-01

    The present report deals with the results of the second phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. The following ranges of variables were studied and 809 burnout measurements were obtained. Pressure 5. 3 2 ; Inlet subcooling 56 sub BO 2 ; Mass velocity 100 2 s; Heated length 600 BO , were plotted against the pressure with the surface heat flux as parameter. The data have been correlated by curves, and the scatter around the curves is less than ± 5 per cent. In the ranges investigated, the observed steam quality at burnout, X BO generally decreases with increasing heat flux and mass velocity but increases with increasing pressure. The data have been compared with the empirical correlation by Tong, and excellent agreement was found for pressures higher than 10 kg/cm 2

  9. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

    International Nuclear Information System (INIS)

    Becker, Kurt M.

    1962-01-01

    The present report deals with the results of the first phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. Data were obtained in the following ranges of variables. Pressure 2.4 sub 2 ; Mass velocity 144 2 /s; Heated length 1040 BO , were plotted against the pressure with the surface heat flux as parameter. The data have been correlated by curves. The scatter of the data around the curves is less than ± 5 per cent. In the ranges investigated the observed steam quality at burnout, x BO generally decreases with increasing heat flux; increases with increasing pressure and decreases with increasing mass velocity. The mass velocity effect has been explained on the basis of climbing film flow theory. Finally we have found that for engineering purposes the effects of inlet subcooling and channel length are negligible

  10. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  11. A study of implementing In-Cycle-Shuffle strategy to a decommissioning boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chung-Yuan, E-mail: tuckjason@iner.gov.tw; Tung, Wu-Hsiung; Yaur, Shyun-Jung

    2017-06-15

    Highlights: • A loading pattern strategy ICS (In-Cycle-Shuffle) was implemented to the last cycle of the boiling water reactor. • The best power sharing distribution and ICS timing was found. • A new parameter “Burnup sharing” is presented to evaluate ICS strategy. - Abstract: In this paper, a loading pattern strategy In-Cycle-Shuffle (ICS) is implemented to the last cycle of the boiling water reactor (BWR) before decommissioning to save the fuel cycle cost. This method needs a core shutdown during the operation of a cycle to change the loading pattern to gain more reactivity. The reactivity model is used to model the ICS strategy in order to find out the best ICS timing and the optimum power sharing distribution before ICS and after ICS. Several parameters of reactivity model are modified and the effect of burnable poison, gadolinium (Gd), is considered in this research. Three cases are presented and it is found that the best ICS timing is at about two-thirds of total cycle length no matter the poisoning effect of Gd is considered or not. According to the optimum power sharing distribution result, it is suggested to decrease the once burnt power and increase the thrice burnt fuel power as much as possible before ICS. After ICS, it is suggested to increase the positive reactivity fuel power and decrease the thrice burnt fuel power as much as possible. A new parameter “Burnup sharing” is presented to evaluate the special case whose EOC power weighting factor and the burnup accumulation factor in the reactivity model are quite different.

  12. A study of implementing In-Cycle-Shuffle strategy to a decommissioning boiling water reactor

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shyun-Jung

    2017-01-01

    Highlights: • A loading pattern strategy ICS (In-Cycle-Shuffle) was implemented to the last cycle of the boiling water reactor. • The best power sharing distribution and ICS timing was found. • A new parameter “Burnup sharing” is presented to evaluate ICS strategy. - Abstract: In this paper, a loading pattern strategy In-Cycle-Shuffle (ICS) is implemented to the last cycle of the boiling water reactor (BWR) before decommissioning to save the fuel cycle cost. This method needs a core shutdown during the operation of a cycle to change the loading pattern to gain more reactivity. The reactivity model is used to model the ICS strategy in order to find out the best ICS timing and the optimum power sharing distribution before ICS and after ICS. Several parameters of reactivity model are modified and the effect of burnable poison, gadolinium (Gd), is considered in this research. Three cases are presented and it is found that the best ICS timing is at about two-thirds of total cycle length no matter the poisoning effect of Gd is considered or not. According to the optimum power sharing distribution result, it is suggested to decrease the once burnt power and increase the thrice burnt fuel power as much as possible before ICS. After ICS, it is suggested to increase the positive reactivity fuel power and decrease the thrice burnt fuel power as much as possible. A new parameter “Burnup sharing” is presented to evaluate the special case whose EOC power weighting factor and the burnup accumulation factor in the reactivity model are quite different.

  13. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  14. On the rationality of the German nuclear phase-out

    International Nuclear Information System (INIS)

    Stoll, Wolfgang

    2018-01-01

    Our state of mind appears to be in equilibrium when it is balanced between opportunity and risk. The relationship between individual expectations of happiness and risk endured varies greatly depending on the state of mind of the individual. It is our understanding of ourselves that manageable individual risks are more likely to be taken than risks imposed by external forces. The anti-nuclear protesters operate skillfully with this super-extension of the term to create general anxiety. However, the problem is of a general nature. Classical scientific findings come mainly from the field of very high probability, which we simply describe as the causal link between cause and effect. In general, however, in the advance of our knowledge into ever more complicated contexts, right down to the so-called statistical ''noise'', the connection between cause and effect is becoming less and less clear. This vagueness opens up a great deal of discretion.

  15. Effects of torus wall flexibility on forces in the Mark I Boiling Water Reactor Pressure Suppression System. Part I

    International Nuclear Information System (INIS)

    Martin, R.W.; McCauley, E.W.

    1977-09-01

    The authors investigated the effects of torus wall flexibility in the pressure suppression system of a Mark I boiling water reactor (BWR) when the torus wall is subjected to hydrodynamic loadings. Using hypothetical models, they examined these flexibility effects under two hydrodynamic loading conditions: (1) a steam relief valve (SRV) discharge pulse, and (2) a loss-of-coolant accident (LOCA) chugging pulse. In the analyses of these events they used a recently developed two-dimensional finite element computer code. Taking the basic geometry and dimensions of the Monticello Mark I BWR nuclear power plant (in Monticello, Minnesota, U.S.A.), they assessed the effects of flexibility in the torus wall by changing values of the inside-diameter-to-wall-thickness ratio. Varying the torus wall thickness (t) with respect to the inside diameter (D) of the torus, they assigned values to the ratio D/t ranging from 0 (infinitely rigid) to 600 (highly flexible). In the case of a modeled steam relief valve (SRV) discharge pulse, they found the peak vertical reaction force on the torus was reduced from that of a rigid wall response by a factor of 3 for the most highly flexible, plant-simulated wall (D/t = 600). The reduction factor for a modeled loss-of-coolant accident (LOCA) chugging pulse was shown to be 1.5. The two-dimensional analyses employed overestimate these reduction factors but have provided, as intended, definition of the effect of torus boundary stiffness. In the work planned for FY79, improved modeling of the structure and of the source is expected to result in factors more directly applicable to actual pressure suppression systems

  16. Application of reliability techniques to prioritize BWR [boiling water reactor] recirculation loop welds for in-service inspection

    International Nuclear Information System (INIS)

    Holman, G.S.

    1989-12-01

    In January 1988 the US Nuclear Regulatory Commission issued Generic Letter 88-01 together with NUREG-0313, Revision 2, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' to implement NRC long-range plans for addressing the problem of stress corrosion cracking in boiling water reactor piping. NUREG-0313 presents guidelines for categorizing BWR pipe welds according to their SCC condition (e.g., presence of known cracks, implementation of measures for mitigating SCC) as well as recommended inspection schedules (e.g., percentage of welds inspected, inspection frequency) for each weld category. NUREG-0313 does not, however, specify individual welds to be inspected. To address this issue, the Lawrence Livermore National Laboratory developed two recommended inspection samples for welds in a typical BWR recirculation loop. Using a probabilistic fracture mechanics model, LLNL prioritized loop welds on the basis of estimated leak probabilities. The results of this evaluation indicate that riser welds and bypass welds should be given priority attention over other welds. Larger-diameter welds as a group can be considered of secondary importance compared to riser and bypass welds. A ''blind'' comparison between the probability-based inspection samples and data from actual field inspections indicated that the probabilistic analysis generally captured the welds which the field inspections identified as warranting repair or replacement. Discrepancies between the field data and the analytic results can likely be attributed to simplifying assumptions made in the analysis. The overall agreement between analysis and field experience suggests that reliability techniques -- when combined with historical experience -- represent a sound technical basis on which to define meaningful weld inspection programs. 13 refs., 8 figs., 5 tabs

  17. Dictionary of nuclear engineering. English-German-French-Russian. Kerntechnik. Englisch-Deutsch-Franzoesisch-Russisch

    Energy Technology Data Exchange (ETDEWEB)

    Sube, R

    1985-01-01

    This volume contains nearly 30,000 entries from all branches of nuclear engineering excluding biological aspects, health physics, and isotope applications in medicine and agriculture. The terminology used in the International Nuclear Information System (INIS) of the IAEA has been completely taken into account. Each English entry is numbered and is followed by corresponding terms in the other languages. Alphabetical indexes of the German, French, and Russian terms are included. At the end of the volume, there is a section devoted to the transliteration of place-names between the English names and their Cyrillic equivalents, and vice versa.

  18. Dealing with the Y2K problem in German nuclear facilities

    International Nuclear Information System (INIS)

    Hagemann, A.

    1999-01-01

    General situation concerning Y2K problem related to german nuclear facilities is presented. Nuclear material used i Germany is owned by EURATOM and Germany is responsible to EURATOM as well as IAEA inspections. Systems of concern are monitoring and control systems, safety related systems and physical protection systems. Present situation is as follows: responsible project teams are formed, Y2K sensitive equipment is identified, designers are contacted, compliance tests specified and schedule of the proof established as of end of August 1999. Experiences obtained in overcoming the Y2K risks are cited

  19. www.kernenergie.de - nuclear power has a German Internet address

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    www.kernenergie.de is the address on the worldwide web under which the German nuclear organizations, Deutsches Atomforum (DAtF), Informationskreis Kernenergie (IK), and Kerntechnische Gesellschaft (KTG) as well as atw - internationale Zeitschrift fuer Kernenergie, and INFORUM can be reached. Extensive sources of information, discussions, on-line dictionaries, computer codes, dynamic web pages, digital documents and multimedia offerings can be called up via the portal under the individual web sites. In this way, www.kernenergie.de provides a comprehensive and up-to-date background of information about nuclear power and adjacent topics in the digital worldwide web. (orig.) [de

  20. Effects of European energy policy on German nuclear phase-out policy

    International Nuclear Information System (INIS)

    Buedenbender, Martin

    2009-01-01

    After the election of the new German government on September 27th, 2009, the nuclear power phase-out decision appears back on top of the political agenda. Hence, an up to date survey of all relevant arguments seems absolutely necessary. In that matter, the scope should not remain national but should also take the European dimension into account. On the European level, a position in favour of nuclear power becomes apparent. Recent political decisions among the 27 member states show a renaissance of atomic energy. EU-Parliament, EU-Commission and EU-Council have all voted for the extensive, long term use of nuclear power in Europe. With its phase-out decision still valid, Germany is part of a minority in Europe. Germany is part of a European market for electricity whose national barriers will blur more and more in the future to form a fully integrated pan-European market in the end. Since nuclear power will provide a major share of the European electricity generation mix, Germany will always be supplied with atomic energy in the long term. This is imperative, regardless of nuclear power plants operating within the borders of Germany or not. Shutting down these facilities in Germany will hence not make the risks associated with atomic energy disappear. It will only add energy-technical challenges to assure long-term supply security. Thus, the new German government should withdraw the phase-out decision. (orig.)

  1. 75 FR 26967 - Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

    Science.gov (United States)

    2010-05-13

    ... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket No. FDA-2010-D-0236] Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory; Availability AGENCY: Food and Drug Administration, HHS. ACTION: Notice. SUMMARY: The Food and Drug...

  2. Steady state thermal hydraulic analysis of a boiling water reactor core, for various power distributions, using computer code THABNA

    International Nuclear Information System (INIS)

    Venkat Raj, V.; Saha, D.

    1976-01-01

    The core of a boiling water reactor may see different power distributions during its operational life. How some of the typical power distributions affect some of the thermal hydraulic parameters such as pressure drop minimum critical heat flux ratio, void distribution etc. has been studied using computer code THABNA. The effect of an increase in the leakage flow has also been analysed. (author)

  3. German writers and the Intermediate-range Nuclear Forces debate in the 1980s

    International Nuclear Information System (INIS)

    Stokes, A.M.

    1991-01-01

    In 1979, NATO announced its decision to deploy American intermediate-range nuclear missiles throughout Western Europe. From then until 1987, when the historic Intermediate-range Nuclear forces (INF) treaty provided for the withdrawal of these weapons as well as those deployed by the Soviets in Eastern Europe, the issue of nuclear weapons preoccupied many Europeans, particularly Germans. Beginning in 1980, fear of nuclear war, with the two Germanies as a potential battlefield, mobilized the largest peace movement that the Federal Republic had witnessed since the fifties, occasioned a massive increase in peace propaganda in East Germany, and brought to public notice that country's first unofficial peace movement. Throughout most of the eighties, writers in both German states opposed missile deployment. This study examines their aims and achievements in this effort and investigates the implications of political engagement for the aesthetic production of selected authors. Analysis of press reports, writers' speeches, interviews, essays and literary texts yielded the following results: INF deployment motivated writers of all political persuasions to take up a variety of peace-oriented pursuits

  4. SWR 1000: an advanced boiling water reactor with passive safety features

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

  5. EXTREMUS - the German solution for act of terrorism - non-nuclear risks coverage

    International Nuclear Information System (INIS)

    Harbrucker, D.

    2004-01-01

    As a consequence of the terrorist attacks of September 11, 2001 and the withdraw of reinsurance capacity worldwide for terrorism coverage under material damage policies EXTREMUS was founded on September 2nd, 2002 by the German insurance industry (shareholders are 16 (re) insurance groups operating in the country). EXTREMUS offers a maximum limit per insured of Euro 1,5 bn. and enjoys a warranty of the German Government to cover accumulation losses in xs of Euro 2.0 bn. up to Euro 10 bn. EXTREMUS only intervenes for policies exceeding Euro 25 mio. due to a self obligation of the primary market to maintain full coverage for smaller risks. Due to the demand of reinsurers losses caused by a b c weapons are excluded. Coverage is not available for nuclear power plants. The paper aims to deals with these issues in more details.(author)

  6. Availability of nuclear fuels: one aspect of the reliability of supply. [German Federal Republic

    Energy Technology Data Exchange (ETDEWEB)

    Dolinski, U; Ziesing, H J [Deutsches Inst. fuer Wirtschaftsforschung, Berlin (F.R. Germany)

    1976-09-01

    In connection with the future supply of nuclear fuels to the Federal Republic of Germany the authors discuss the problems which arise. They describe the future development of the demand for natural uranium and the work involved in separation and deal with the possibilities of meeting the demand. They pay particular attention to the changed market situation and to the policies of the countries which produce uranium. The article is a detailed examination of the report by the German Institute for Economic Research (DIW) on aspects of the reliability of supply to the electricity supply industry in the Federal Republic of Germany, which was prepared under the aegis of the Bavarian Ministry for Economic and Transport. This investigation will be published, omitting the regional aspects, towards the middle of 1976, by the German Institute for Economic Research under the title 'Reliability, price and environmental protection aspects of energy supply'.

  7. Advanced handbook for accident analyses of German nuclear power plants; Weiterentwicklung eines Handbuches fuer Stoerfallanalysen deutscher Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Kerner, Alexander; Broecker, Annette; Hartung, Juergen; Mayer, Gerhard; Pallas Moner, Guim

    2014-09-15

    The advanced handbook of safety analyses (HSA) comprises a comprehensive electronic collection of knowledge for the compilation and conduction of safety analyses in the area of reactor, plant and containment behaviour as well as results of existing safety analyses (performed by GRS in the past) with characteristic specifications and further background information. In addition, know-how from the analysis software development and validation process is presented and relevant rules and regulations with regard to safety demonstration are provided. The HSA comprehensively covers the topic thermo-hydraulic safety analyses (except natural hazards, man-made hazards and malicious acts) for German pressurized and boiling water reactors for power and non-power operational states. In principle, the structure of the HSA-content represents the analytical approach utilized by safety analyses and applying the knowledge from safety analyses to technical support services. On the basis of a multilevel preparation of information to the topics ''compilation of safety analyses'', ''compilation of data bases'', ''assessment of safety analyses'', ''performed safety analyses'', ''rules and regulation'' and ''ATHLET-validation'' the HSA addresses users with different background, allowing them to enter the HSA at different levels. Moreover, the HSA serves as a reference book, which is designed future-oriented, freely configurable related to the content, completely integrated into the GRS internal portal and prepared to be used by a growing user group.

  8. Experience in commissioning and scientific support of nuclear power plant operation in the German Democratic Republic

    International Nuclear Information System (INIS)

    Ackermann, G.; Endler, A.; Loth, K.H.

    1983-01-01

    In the German Democratic Republic two nuclear power plants equipped with four pressurized-water WWER-440-type reactors were commissioned in the years 1973-1979 and have since been in routine operation. The nuclear power plants (NPPs) in the GDR were constructed on the basis of projects developed and tested in the Soviet Union. Their main equipment was also supplied by the USSR. Under the technical supervision of USSR scientific institutions and design organizations, the construction, commissioning and operation of NPPs had to be prepared and organized by the design organizations, scientific institutions and the prospective user in the German Democratic Republic. To this effect, and to ensure nuclear safety, availability and maximum exploitation of nuclear fuel, it was necessary to develop national scientific capacities. The organizations of the USSR supply basic and supplementary documents prescribing the required extent of investigations to prove that the equipment and systems will meet the projected and operational conditions from loading to commercial operation. Concrete test programmes for commissioning and operational instructions are elaborated by national organizations. Pre-operational and initial startup tests are performed by the personnel of the plant, of enterprises and scientific institutions under the technical supervision of the design organizations of the USSR and the German Democratic Republic. Proceeding from experience gained for more than 15 years in the scientific support of commissioning and operation, the required training and disciplines of personnel are reported. The tasks to be met by the various disciplines are described. Furthermore, information is given on the main stages of commissioning and on a number of results. Finally, selected results of work in connection with operational assistance are reported

  9. The Franko-German aproach for a nuclear power plant in Turkey

    International Nuclear Information System (INIS)

    Ruess, F.

    1994-01-01

    Nuclear Power International (N P1) has been established as a joint subsidiary of Siemens and Framatome in 1989, thereby combining the experience accumulated in both parent companies with more than 100,000 MW capacity installed or on order in nuclear field. We intend to compete in a potential nuclear power project in Turkey on the basis of the German Pressurized Water Reactor Technology. We intend to establish a Consortium which on the foreign suppliers side will include Siemens, Framatome and GEC-Alsthom. In addition to the foreign partners in the Consortium we will include the Turkish industry in our proposal in order to achieve a maximum possible local content, which in our previous proposal was in the range of 30 % of the contract-value

  10. The rise and the crisis of the German nuclear economy 1945-1975

    International Nuclear Information System (INIS)

    Radkau, J.

    1983-01-01

    The nuclear economy has for some time been complaining of being the scapegoat of a protest movement the origin and motive of which has nothing to do with nuclear technology. On the opposite side the sticker ''Nuclear energy - No thanks'' has long become the distinguishing sign of people longing to change their entire way of life. The fact that this protest movement around 1970 did spring from blatantly weak points of the nuclear technology which at the time became obvious has practically fallen into oblivion. The author describes the history of the German nuclear economy under changing aspects in its sociopolitical, economic and technical context. Files of the ministries in charge and their advisory committees, minites of the committees of the Bundestag, bequests of influential personages and oral communications - material which for the most part has not so far been either used or published - provide insight into connections unexplained until now or unknown altogether. This historic recourse might lead to the subject as such, viz. nuclear technology, becoming again the object of discussions instead of ''Nuclear energy. No thanks'' being used as a sticker. (orig./HP) [de

  11. Development and construction of nuclear power and nuclear heating stations in the USSR

    International Nuclear Information System (INIS)

    Schmidt, G.; Kirmse, B.

    1983-01-01

    The state-of-the-art of nuclear power technology in the USSR is reviewed by presenting characteristic data on design and construction. The review takes into consideration the following types of facilities: Nuclear power stations with 1000 MWe pressurized water reactors, with 1000 MWe pressure tube boiling water reactors, and with 600 MWe fast breeder reactors; nuclear heating power stations with 1000 MWe reactors and nuclear heating stations with 500 MWth boiling water reactors

  12. German experience in recycling of ferrous metallic residues from nuclear decommissioning by melting

    International Nuclear Information System (INIS)

    Quade, U.; Kluth, Th.

    2008-01-01

    Due to the delay of commissioning of final depositories for nuclear waste on the one hand and the increasing amount of steel scrap resulting from operation and decommissioning of nuclear facilities on the other hand, recycling of ferrous metal scrap to packagings made of ductile cast iron becomes more and more economical. A pool of know-how from waste managers, radiation protection experts, metallurgists and foundry experts and their teamwork is required to run this recycling path successfully. Siempelkamp provides this combination of experience by operating a melting facility for slightly radioactive contaminated scrap as well as a foundry for manufacturing of ductile cast iron products for the nuclear industry, both licensed by the German Radiation Protection Ordinance. In 1989, the CARLA plant (Centrale Anlage zum Rezyklieren von leichtradioaktiven Abfollen) started operation. A medium frequency induction furnace with a capacity of 3,2 t is core of the plant. Tools for dismantling and cutting components to chargeable sizes are available. From the total of 23000 t of melted scrap, 12000 t have been recycled to the manufacturing of containers for transport and storage of medium- and high active waste and for shielding plates. Manufacture of the castings takes place in the Siempelkamp foundry located at the same site. 8000 t of melted scrap could be released for industrial recycling. Scrap metal which does not meet the metallurgical specification for cast iron, is converted into iron granules. Up to now more than 2000 t of iron granules have been recycled as additive for heavy concrete containers. This production is in cooperation with an external partner. With regard to the German situation, the cost for recycling is only half compared to high pressure compaction, long-term interim storage and final disposal. The advantage of recycling is approx. 90 % less volume compared to the volume resulting from other disposal paths. It can be concluded that the German

  13. Development of the on-site power supply in German nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Simon, M. [Gesellschaft fuer Reaktorsicherheit - GRS mbH, Schwertnergasse 1, D-5000 Koeln 1, Cologne (Germany)

    1986-02-15

    The design of the on-site power supply is different in German Nuclear Power Plants, depending on age and size of the plant. The cause for this is the evolution of the safety requirements. The general development of the design of safety Systems, which resulted in a strict separation of redundant trains is also reflected in the design of the emergency power system and even the complete on-site power supply System. This will be demonstrated by different examples. The advantages of this design with respect to the availability of on-site power will be explained and verified by means of operating experience. (author)

  14. Availability of nuclear fuels: an aspect of supply assurance. [German Federal Republic

    Energy Technology Data Exchange (ETDEWEB)

    Dolinski, U; Ziesing, H J [Deutsches Inst. fuer Wirtschaftsforschung, Berlin (F.R. Germany)

    1976-08-01

    The future supply possibilities for the Federal Republic of Germany with nuclear fuel are investigated by the German Institute for Economy Research in an expertise on 'the safety, price and environmental aspects in the energy supply'. The requirements of natural uranium in the Federal Republic of Germany are dealt with, as well as the world-wide assured and probable deposits of natural uranium in relation to the extraction costs. After indicating the restrictive changes in the export policy of the supplying countries, the international and national state of development of recycling is shown.

  15. Case studies on the feasibility of the transient analysis system STAR in German nuclear power plants

    International Nuclear Information System (INIS)

    Buettner, W.E.; Felkel, L.; Zapp, A.

    1984-01-01

    On the basis of distubances which actually have occurred in German nuclear power plants a case-study has been performed to evaluate the feasibility of the computer-based disturbance analysis system STAR. By means of a compact plant simulator the disturbances collected have been remodelled and anlysed, on-line, with the disturbance analysis system STAR. In the last phase of the project experiments have been performed with reactor operators to get their reaction to, and opinion on, computerbased-operator aids. (orig.) [de

  16. Franco-German nuclear cooperation: from the 'common product' to the first European pressurized water reactor

    International Nuclear Information System (INIS)

    Vignon, D.

    1999-01-01

    It has now been 10 years since Framatome and Siemens decided to collaborate on the design and sales of an advanced nuclear power plant (NPP) model based on pressurized water reactor (PWR) technology. Originally called the 'common product', this model was renamed the European pressurized water reactor when Electricite de France (EDF) and the German electric utilities joined this cooperative development effort in 1992. Since the beginning, this cooperation has been formalized in the framework of an agreement that led to the founding of a joint and equally owned subsidiary, Nucler Power International (NPI), which is reponsible for leading the development of the new model and later handling its export sales

  17. Feasibility of core management system by data communication for boiling water reactors

    International Nuclear Information System (INIS)

    Motoda, H.; Tanisaka, S.; Kiguchi, T.; Yonenaga, H.

    1977-01-01

    A core management system by data communication has been designed and proposed for more efficient operation of boiling water reactor (BWR) plants by faster transmission and centralized management of information. The system comprises three kinds f computers: a process computer for monitoring purposes at the reactor site, a center computer for administration purposes at the head office, and a large scientific computer for planning and evaluation purposes. The process and the large computers are connected to the center computer by a data transmission line. To demonstrate the feasibility of such a system, the operating history evaluation system, which is one of the subsystems of the core management system, has been developed along the above concept. Application to the evaluation of the operating history of a commercial BWR shows a great deal of merit. Quick response and a significant manpower reduction can be expected by data communication and minimized intervention of human labor. Visual display is also found to be very useful in understanding the core characteristics

  18. TARMS, an on-line boiling water reactor operation management system

    International Nuclear Information System (INIS)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-01-01

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.) [de

  19. Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

    International Nuclear Information System (INIS)

    Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

    2008-01-01

    In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

  20. Development of a real-time stability measurement system for boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; King, W.T.

    1987-01-01

    This paper describes the development of a portable, real time system for boiling water reactor (BWR) stability measurements. The system provides a means for the operator to monitor the reactor stability using existing plant instrumentation and commercially available hardware. The noise component (i.e., perturbations around steady state) of the neutron signal in BWRs has been shown to contain information about reactor stability, and several algorithms have been developed to extract that information. For the present work, the authors have used an algorithm that has been implemented on a portable personal computer. This algorithm uses the autocorrelation function of naturally occurring neutron noise (measured without special plant perturbations) and an autoregressive modeling technique to produce the asymptotic DR. For this real-time implementation, neutron noise data is preconditioned (i.e., filtered and amplified) and sampled at a 5-Hz sampling rate using a commercial data-acquisition system. Approximately every 1.5 min, the current (snapshot) autocorrelation function is computed directly from the data, and the average autocorrelation is updated. The current and average DR estimates are evaluated with the same periodicity and are displayed on the screen along with the autocorrelations and average power spectrum of the neutron noise

  1. Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2002-01-01

    A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies such as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained

  2. Artificial neural networks versus conventional methods for boiling water reactor stability monitoring

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der

    1995-01-01

    The application of an artificial neural network (ANN) for boiling water reactor (BWR) stability monitoring was studied. A three-layer perceptron was trained on synthetic autocorrelation functions to estimate the decay ratio and the resonance frequency from measured neutron noise. Training of the ANN was improved by adding noise to the training patterns and by applying nonconventional error definitions in the generalized delta rule. The performance of the developed ANN was compared with those of conventional stability monitoring techniques. Explicit care was taken for generating unbiased test data. It is found that the trained ANN is capable of monitoring the stability of the Dodewaard BWR for four specific cases. By comparing properties such as the false alarm ratio, the alarm failure ratio, and the average time to alarm, it is shown that it performs worse than model-based methods in stability monitoring of exact second-order systems but that it is more robust (better resistant to corruptions of the input data and to deviations of the system at issue from an exact second-order system) than other methods. The latter explains its good performance on the Dodewaard BWR and is promising for the application of an ANN for stability monitoring of other reactors and for other operating conditions

  3. Measurement on the effect of sound wave in upper plenum of boiling water reactor

    International Nuclear Information System (INIS)

    Kumagai, Kosuke; Someya, Satoshi; Okamoto, Koji

    2009-01-01

    In recent years, the power uprate of Boiling Water Reactors have been conducted at several existing power plants as a way to improve plant economy. In one of the power uprated plants (117.8% uprates) in the United States, the steam dryer breakages due to fatigue fracture occurred. It is conceivable that the increased steam flow passing through the branches caused a self-induced vibration with the propagation of sound wave into the steam-dome. The resonance among the structure, flow and the pressure fluctuation resulted in the breakages. To understand the basic mechanism of the resonance, previous researches were done by a point measurement of the pressure and by a phase averaged measurement of the flow, while it was difficult to detect the interaction among them by the conventional method. In this study, Dynamic Particle Image Velocimetry (PIV) System was applied to investigate the effect of sound on natural convection and forced convection. Especially, when the phases of acoustic sources were different, various acoustic wave effects were checked. (author)

  4. Investigation and examination on the cracking of pipings in boiling water reactors

    International Nuclear Information System (INIS)

    1977-01-01

    This is the report made by the Reactor Safety Technology Expert Committee to the Atomic Energy Commission regarding the investigation and examination on stress corrosion cracking which seems to be the cause of the cracking of pipings in boiling water reactors, the measures to reduce it, and the subjects of research hereafter. Recently, the stress corrosion cracking of primary coolant pipings has been often observed, and this phenomenon occurred in the pressure boundary of primary coolant, consequently it is possible to be linked to the troubles of large scale. The Reactor Material Subcommittee was established on May 14, 1975, and investigated the cracking phenomena in the recirculating system and core spray system of BWRs in Japan and foreign countries. The recent cases have been concentrated to the heat-affected part due to welding of 304 type austenitic stainless steel pipings of from 4 in to 10 in diameter for BWRs. They are the stress corrosion cracking at grain boundaries occurred under the loaded condition and in the environment of high temperature, high pressure water. The cracking of this kind was never experienced in PWRs. The results of the technical examination, the consideration of the mechanism of stress corrosion cracking, and the countermeasures are described. (Kako, I.)

  5. Influence of subcooled boiling on out-of-phase oscillations in boiling water reactors

    International Nuclear Information System (INIS)

    Munoz-Cobo, J.L.; Chiva, S.; Escriva, A.

    2005-01-01

    In this paper, we develop a reduced order model with modal kinetics for the study of the dynamic behavior of boiling water reactors. This model includes the subcooled boiling in the lower part of the reactor channels. New additional equations have been obtained for the following dynamics magnitudes: the effective inception length for subcooled boiling, the average void fraction in the subcooled boiling region, the average void fraction in the bulk-boiling region, the mass fluxes at the boiling boundary and the channel exit, respectively, and so on. Each channel has three nodes, one of liquid, one with subcooled boiling, and one with bulk boiling. The reduced order model includes also a modal kinetics with the fundamental mode and the first subcritical one, and two channels representing both halves of the reactor core. Also, in this paper, we perform a detailed study of the way to calculate the feedback reactivity parameters. The model displays out-of-phase oscillations when enough feedback gain is provided. The feedback gain that is necessary to self-sustain these oscillations is approximately one-half the gain that is needed when the subcooled boiling node is not included

  6. Optimization of fuel exchange machine operation for boiling water reactors using an artificial intelligence technique

    International Nuclear Information System (INIS)

    Sekimizu, K.; Araki, T.; Tatemichi, S.I.

    1987-01-01

    Optimization of fuel assembly exchange machine movements during periodic refueling outage is discussed. The fuel assembly movements during a fuel shuffling were examined, and it was found that the fuel assembly movements consist of two different movement sequences;one is the ''PATH,'' which begins at a discharged fuel assembly and terminates at a fresh fuel assembly, and the other is the ''LOOP,'' where fuel assemblies circulate in the core. It is also shown that fuel-loading patterns during the fuel shuffling can be expressed by the state of each PATH, which is the number of elements already accomplished in the PATH actions. Based on this fact, a scheme to determine a fuel assembly movement sequence within the constraint was formulated using the artificial intelligence language PROLOG. An additional merit to the scheme is that it can simultaneously evaluate fuel assembly movement, due to the control rods and local power range monitor exchange, in addition to normal fuel shuffling. Fuel assembly movements, for fuel shuffling in a 540-MW(electric) boiling water reactor power plant, were calculated by this scheme. It is also shown that the true optimization to minimize the fuel exchange machine movements would be costly to obtain due to the number of alternatives that would need to be evaluated. However, a method to obtain a quasi-optimum solution is suggested

  7. Expert system for maintenance management of a boiling water reactor power plant

    International Nuclear Information System (INIS)

    Hong Shen; Liou, L.W.; Levine, S.; Ray, A.; Detamore, M.

    1992-01-01

    An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, which must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components

  8. Nonlinear dynamics and stability of boiling water reactors: qualitative and quantitative analyses

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1985-01-01

    A phenomenological model has been developed to simulate the qualitative behavior of boiling water reactors (BWRs) in the nonlinear regime under deterministic and stochastic excitations. After the linear stability threshold is crossed, limit cycle oscillations appear due to interactions between two unstable equilibrium points and the phase-space trajectories. This limit cycle becomes unstable when the feedback gain exceeds a certain critical value. Subsequent limit cycle instabilities produce a cascade of period-doubling bifurcations that leads to a periodic pulsed behavior. Under stochastic excitations, BWRs exhibit a single characteristic resonance, at approx.0.5 Hz, in the linear regime. By contrast, this work shows that harmonics of this characteristic frequency appear in the nonlinear regime. Furthermore, this work also demonstrates that amplitudes of the limit cycle oscillations do not depend on the variance of the stochastic excitation and remain bounded at all times. A physical model of nonlinear BWR dynamics has also been developed and employed to calculate the amplitude of limit cycle oscillations and their effects on fuel integrity over a wide range of operating conditions in the Vermont Yankee reactor. These calculations have confirmed that, beyond the threshold for linear stability, the reactor's state variable undergo limit cycle oscillations

  9. Status report on the Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D ampersand D) Project

    International Nuclear Information System (INIS)

    Sears, L.; Garlock, G.; Mencarelli, R.; Fellhauer, C.

    1994-01-01

    ALARON Corporation is under contract, to Argonne National Laboratory - East (ANL-E), to complete the decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR). The project, begun, in 1986 by ANL-E personnel, is projected to be completed by the end of 1994. The final phase of work was awarded to ALARON in December 1993 with the scope of work including the disassembly and removal of all remaining reactor internals, the reactor vessel, the lead bio-shield, the core liner, and the activated portion of the concrete bio-shield. This paper discusses the work undertaken beginning in January 1994 and continuing through July 1994. During this period the required pre-mobilization documentation was prepared and approved, mobilization was completed, and the reactor internals, reactor vessel, lead bio-shield and core liner were removed. The paper will compare the planned schedule to the actual schedule, discuss problems encountered, review volume reduction techniques and health and safety issues including radiological aspects of the project

  10. Data Reconciliation in the Steam-Turbine Cycle of a Boiling Water Reactor

    International Nuclear Information System (INIS)

    Sunde, Svein; Berg, Oivind; Dahlberg, Lennart; Fridqvist, Nils-Olof

    2003-01-01

    A mathematical model for a boiling water reactor steam-turbine cycle was assembled by means of a configurable, steady-state modeling tool TEMPO. The model was connected to live plant data and intermittently fitted to these by minimization of a weighted least-squares object function. The improvement in precision achieved by this reconciliation was assessed from quantities calculated from the model equations linearized around the minimum and from Monte Carlo simulations. It was found that the inclusion of the flow-passing characteristics of the turbines in the model equations significantly improved the precision as compared to simple mass and energy balances, whereas heat transfer calculations in feedwater heaters did not. Under the assumption of linear model equations, the quality of the fit can also be expressed as a goodness-of-fit Q. Typical values for Q were in the order of 0.9. For a validated model Q may be used as a fault detection indicator, and Q dropped to very low values in known cases of disagreement between the model and the plant state. The sensitivity of Q toward measurement faults is discussed in relation to redundancy. The results of the linearized theory and Monte Carlo simulations differed somewhat, and if a more accurate analysis is required, this is better based on the latter. In practical application of the presently employed techniques, however, assessment of uncertainties in raw data is an important prerequisite

  11. Fact and fiction in ECP measurement and control in boiling water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2005-01-01

    A review is presented of various electrochemical potentials, including the electrochemical corrosion potential (ECP), that are used in the mitigation of stress corrosion cracking in the primary coolant circuits of boiling water reactors (BWRs). Attention is paid to carefully defining each potential in terms of fundamental electrochemical concepts, so as to counter the confusion that has arisen due to the misuse of previously accepted terminology. A brief discussion is also included of reference electrodes and it is shown on the basis of experimental data that the use of a platinum redox sensor as a reference electrode in the monitoring of ECP in BWR primary coolant circuits is inappropriate and should be discouraged. If platinum is used as a reference electrode, because of extenuating circumstances (e.g., potential measurements in high dose regions in a reactor core), the onus must be placed on the user to demonstrate quantitatively that the electrode behaves as an equilibrium electrode under the specified conditions and/or that its potential is invariant with changes in the independent variables of the system. Preferably, a means should also be demonstrated of transferring the measured potential to the standard hydrogen electrode (SHE) scale. (orig.)

  12. Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

    2017-05-15

    Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

  13. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts (Part 2)

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Persson, P; Nilsson, L; Eriksson, O

    1963-06-15

    The present report deals with the results of the second phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. The following ranges of variables were studied and 809 burnout measurements were obtained. Pressure 5. 3 < p < 37. 3 kg/cm{sup 2}; Inlet subcooling 56 < {delta}t{sub sub} < 212 deg C; Steam quality 0. 20 < x{sub BO} < 0.95; Heat Flux 50 < q/A < 515 W/cm{sup 2}; Mass velocity 100 < m'/F < 1890 kg/m{sup 2}s; Heated length 600 < L < 2500 mm; Duct diameter d = 10 mm. The results are presented in diagrams, where for a certain geometry, the burnout steam qualities, x{sub BO} , were plotted against the pressure with the surface heat flux as parameter. The data have been correlated by curves, and the scatter around the curves is less than {+-} 5 per cent. In the ranges investigated, the observed steam quality at burnout, X{sub BO} generally decreases with increasing heat flux and mass velocity but increases with increasing pressure. The data have been compared with the empirical correlation by Tong, and excellent agreement was found for pressures higher than 10 kg/cm{sup 2}.

  14. Materials behavior in alternate (hydrogen) water chemistry in the Ringhals-1 boiling water reactor

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Cubicciotti, D.; Trolle, M.

    1986-01-01

    In-plant studies on the intergranular stress corrosion cracking (IGSCC) of sensitized austenitic stainless steel (SS) have been performed at the Swedish Ringhals-1 boiling water reactor (BWR). The studies have covered the present [full-temperature (normal)] water chemistry (PWC) and the alternate (primary) water chemistry (AWC) with hydrogen addition. The test techniques applied were constant extension rate testing (CERT) and electrochemical potential (ECP) measurements. The program was covered by extensive environment monitoring. The results verify earlier laboratory studies which show that sensitized austenitic SS is susceptible to IGSCC in PWC, but not in AWC. Other pressure-bearing BWR construction materials are not adversely affected by AWC. The boundary conditions in Ringhals-1 have been established for an AWC, which is defined as an environment that does not produce IGSCC in sensitized SS. The results are compared with a similar program at Dresden-2, and the points of agreement and discordance in the results are discussed. The relevance of ECP measurements for the control of AWC is discussed

  15. Variations in the chemical speciation behaviour of radioiodines in the Tarapur Boiling Water Reactor

    International Nuclear Information System (INIS)

    Venkateswaran, G.; Gokhale, A.S.; Moorthy, P.N.

    1998-01-01

    The chemical behaviour of radioiodines in the primary coolant of the Tarapur Boiling Water Reactor has been studied under different operating conditions. During normal operation, radioiodines speciated mainly as I - (≅60%) and IO 3 - (≅35%) with 2 . At 1-5 h into reactor shutdown conditions, radioiodines existed predominantly as IO 3 - species (>80%). Beyond 5 h after shutdown, quantitative conversion of IO 3 - to I - was observed to occur in about 20 h duration. Long time after reactor shutdown, radioiodines were present in the coolant as I - species only. A quantitative conversion of near carrier-free IO 3 - to I - was observed in laboratory low dose rate (0.95 kGy/h), low and high dose gamma irradiation experiments in near neutral solutions both in absence and presence of externally added H 2 O 2 . However, near carrier-free I - solutions irradiated under the same conditions yielded ≅15% IO 3 - species only which is in agreement with the literature data. The radioiodine speciation behaviour in reactor water has been explained by a qualitative model coupling iodine release from defective fuel elements and the associated gamma irradiation effects. (author)

  16. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors

    International Nuclear Information System (INIS)

    Fabreca, S.

    1964-10-01

    The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [fr

  17. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts (Part 2)

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Mathisen, R P; Eklind, O; Norman, B

    1964-01-15

    The hydrodynamic stability and the burnout conditions for flow of boiling water have been studied in a natural circulation loop in the pressure range from 10 to 70 atg. The test section was a round, duct of 20 mm inner diameter and 4890 mm heated length. The experimental results showed that within the ranges tested the stability of the flow increases with increasing pressure, increasing throttling before the test section, but decreases with increasing inlet sub-cooling and increasing throttling after the test section. The measured thresholds of instability compared well with the analytical results by Jahnberg. For an inlet sub-cooling temperature of about 2 deg C the measured burnout steam qualities were low by a factor of about 1.3 compared to forced circulation data obtained with the same test section. At higher sub-cooling temperatures the discrepancy between forced and natural circulation data increased, so that at {delta}t{sub sub} = 16 deg C, the natural circulation data were low by a factor of about 2.5. However, by applying inlet throttling of the flow the burnout values approached and finally coincided with the forced circulation data.

  18. U.S. experience with hydrogen water chemistry in boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Head, R.A.; Indig, M.E.; Ruiz, C.P.; Simpson, J.L.

    1988-01-01

    Hydrogen water chemistry in boiling water reactors is currently being adopted by many utilities in the U.S., with eleven units having completed preimplementation test programs, four units operating permanently with hydrogen water chemistry, and six other units in the process of installing permanent equipment. Intergranular stress corrosion cracking protection is required for the recirculation piping system and other regions of the BWR systems. The present paper explores progress in predicting and monitoring hydrogen water chemistry response in these areas. Testing has shown that impurities can play an important role in hydrogen water chemistry. Evaluation of their effects are also performed. Both computer modeling and in plant measurements show that each plant will respond uniquely to feedwater hydrogen addition. Thus, each plant has its own unique hydrogen requirement for recirculation system protecion. Furthermore, the modeling, and plant measurements show that different regions of the BWR respond differently to hydrogen injection. Thus, to insure protection of components other than the recirculation systems may require more (or less) hydrogen demand than indicated by the recirculation system measurements. In addition, impurities such as copper can play a significant role in establishing hydrogen demand. (Nogami, K.)

  19. A two-step method for developing a control rod program for boiling water reactors

    International Nuclear Information System (INIS)

    Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift

  20. A study on boiling water reactor regional stability from the viewpoint of higher harmonics

    International Nuclear Information System (INIS)

    Takeuchi, Yutaka; Takigawa, Yukio; Uematsu, Hitoshi

    1994-01-01

    A quantitative study on a mechanism for boiling water reactor regional stability has been carried out from the viewpoint of higher harmonics. In the mechanism, the gain decrease in the void-to-power transfer function can be explained by the higher harmonics mode subcriticality. It is shown that the thermal-hydraulic feedback effect can compensate for the gain decrease, and regional oscillation can be sustained that way. For quantitative evaluations, a three-dimensional higher harmonics analysis model has been developed. The results show that the first azimuthal harmonics subcriticality has a relatively small value under a regionally unstable condition. Comparing the subcriticality and the steady-state power distribution, it is shown that the distribution exists whose first azimuthal harmonics subcriticality takes a small value. A method of decomposition for the oscillated power responses into the harmonics modes is presented. The results show that the corewide oscillation power response consists almost entirely of the fundamental mode, and the regional oscillation power response consists almost entirely of the first azimuthal harmonics mode. This indicates that regional oscillation is a phenomenon in which the first azimuthal harmonics mode oscillates on the basis of the fundamental mode

  1. Higher order generalized perturbation theory for boiling water reactor in-core fuel management optimization

    International Nuclear Information System (INIS)

    Moore, B.R.; Turinsky, P.J.

    1998-01-01

    Boiling water reactor (BWR) loading pattern assessment requires solving the two-group, nodal form of the neutron diffusion equation and drift-flux form of the fluid equations simultaneously because these equation sets are strongly coupled via nonlinear feedback. To reduce the computational burden associated with the calculation of the core attributes (that is, core eigenvalue and thermal margins) of a perturbed BWR loading pattern, the analytical and numerical aspects of a higher order generalized perturbation theory (GPT) method, which correctly addresses the strong nonlinear feedbacks of two-phase flow, have been established. Inclusion of Jacobian information in the definition of the generalized flux adjoints provides for a rapidly convergent iterative method for solution of the power distribution and eigenvalue of a loading pattern perturbed from a reference state. Results show that the computational speedup of GPT compared with conventional forward solution methods demanding consistent accuracy is highly dependent on the number of spatial nodes utilized by the core simulator, varying from superior to inferior performance as the number of nodes increases

  2. Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Boing, L.E.; Aldana, J.

    1997-03-01

    The Final Report for the Decontamination and Decommissioning (D ampersand D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D ampersand D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D ampersand D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D ampersand D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a open-quotes Radiologically Controlled Area,close quotes noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion)

  3. Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M

    1962-07-01

    The present report deals with the results of the first phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. Data were obtained in the following ranges of variables. Pressure 2.4

  4. Automatic boiling water reactor control rod pattern design using particle swarm optimization algorithm and local search

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Cheng-Der, E-mail: jdwang@iner.gov.tw [Nuclear Engineering Division, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China); Lin, Chaung [National Tsing Hua University, Department of Engineering and System Science, 101, Section 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China)

    2013-02-15

    Highlights: ► The PSO algorithm was adopted to automatically design a BWR CRP. ► The local search procedure was added to improve the result of PSO algorithm. ► The results show that the obtained CRP is the same good as that in the previous work. -- Abstract: This study developed a method for the automatic design of a boiling water reactor (BWR) control rod pattern (CRP) using the particle swarm optimization (PSO) algorithm. The PSO algorithm is more random compared to the rank-based ant system (RAS) that was used to solve the same BWR CRP design problem in the previous work. In addition, the local search procedure was used to make improvements after PSO, by adding the single control rod (CR) effect. The design goal was to obtain the CRP so that the thermal limits and shutdown margin would satisfy the design requirement and the cycle length, which is implicitly controlled by the axial power distribution, would be acceptable. The results showed that the same acceptable CRP found in the previous work could be obtained.

  5. Fuzzy logic control of water level in advanced boiling water reactor

    International Nuclear Information System (INIS)

    Lin, Chaung; Lee, Chi-Szu; Raghavan, R.; Fahrner, D.M.

    1995-01-01

    The feedwater control system in the Advanced Boiling Water Reactor (ABWR) is more challenging to design compared to other control systems in the plant, due to the possible change in level from void collapses and swells during transient events. A basic fuzzy logic controller is developed using a simplified ABWR mathematical model to demonstrate and compare the performance of this controller with a simplified conventional controller. To reduce the design effort, methods are developed to automatically tune the scaling factors and control rules. As a first step in developing the fuzzy controller, a fuzzy controller with a limited number of rules is developed to respond to normal plant transients such as setpoint changes of plant parameters and load demand changes. Various simulations for setpoint and load demand changes of plant performances were conducted to evaluate the modeled fuzzy logic design against the simplified ABWR model control system. The simulation results show that the performance of the fuzzy logic controller is comparable to that of the Proportional-Integral (PI) controller, However, the fuzzy logic controller produced shorter settling time for step setpoint changes compared to the simplified conventional controller

  6. A diagnostic expert system for a boiling water reactor using a dynamic model

    International Nuclear Information System (INIS)

    Sonoda, Y.; Kanemoto, S.; Imaruoka, H.

    1990-01-01

    A diagnostic expert system for abnormal disturbances in a BWR (Boiling Water Reactor) plant has been developed. The peculiar feature of this system is a diagnostic method which combines artificial intelligence technique with numerical analysis technique. The system has three diagnostic functions, 1) identification of anomaly position (device or sensor), 2) identification of anomaly mode and 3) identification of anomaly cause. Function 1) is implemented as follows. First, a hypothesis about anomaly propagation paths is built up by qualitative reasoning, using knowledge of causal relations among observed signals. Next, the abnormal device or sensor is found by applying model reference method and fuzzy set theory to test the hypothesis, using knowledge of plant structure and function, heuristic strategy of diagnosis and module type dynamic simulator. This simulator is composed of basic transfer function modules. The simulation model for the testing region is built up automatically, according to the requirement from the diagnostic task. Function 2) means identification of dynamic characteristics for an anomaly. It is realized by tuning model parameters so as to reproduce the abnormal signal behavior using the non-linear programing method. Function 3) derives probable anomaly causes from heuristic rules between anomaly mode and cause. A basic plant dynamic model was built up and adjusted to dynamic characteristics for one BWR plant (1100MWe). In order to verify the diagnostic functions of this system, data for several abnormal events was compiled by modifying this model. The diagnostic functions were proved useful, through the simulated abnormal data

  7. Water chemistry in boiling water reactors - A Leibstadt-specific overview

    International Nuclear Information System (INIS)

    Sarott, F.-A.

    2005-01-01

    The boiling water reactor (BWR) consists of two main water circuits: the water-steam cycle and the main cooling water system. In the introduction, the goals and tasks of the BWR plant chemistry are described. The most important objectives are the prevention of system degradation by corrosion and the minimisation of radiation fields. Then a short description of the BWR operation principle, including the water steam cycle, the transport of various impurities by the steam, removing impurities from the condensate, the reactor water clean-up system, the balance of plant and the main cooling water system, is given. Subsequently, the focus is set on the water-steam cycle chemistry. In order to fulfil the somewhat contradictory requirements, the chemical parameters must be well balanced. This is achieved by the water chemistry control method called 'normal water chemistry'. Other additional methods are used for the solution to different problems. The 'zinc addition method' is applied to reduce high radiation levels around the recirculation loops. The 'hydrogen water chemistry method' and the 'noble metal chemical addition method' are used to protect the reactor core components and piping made of stainless steel against stress corrosion cracking. This phenomenon has been observed for about 40 years and is partly due to the strong oxidising conditions in the BWR water. Both mitigation methods are used by the majority of the BWR plants all over the world (including the two Swiss NPPs Muehleberg and Leibstadt). (author)

  8. Progressive or accelerated phasing out nuclear? Lessons learned from the German choice

    International Nuclear Information System (INIS)

    Leveque, Francois

    2013-01-01

    This article first recalls the decision taken at the end of the 1990's by the SPD German government to close the last nuclear reactor in 2022. Back to power, the CDU which notably previously announced an extension of nuclear reactor lifetime and a development of renewable energy as an energy transition policy, but faces the Fukushima accident which resulted in a important change and a decision to close nuclear reactors. This decision resulted in an economic loss: cheap kWh's are to be replaced by expensive ones. The author states that phasing out nuclear would cost 63.8 billions euros. He also states that the macro-economic assessment might be negative although macro-economic effects are difficult to be precisely foreseen. Another consequence will be less production of nuclear wastes but more polluting atmospheric emissions. He thinks that a low carbon price will not affect coal-based electricity production. He finally comments the perception of the nuclear risk

  9. Potential uses of high gradient magnetic filtration for high-temperature water purification in boiling water reactors

    International Nuclear Information System (INIS)

    Elliott, H.H.; Holloway, J.H.; Abbott, D.G.

    1979-01-01

    Studies of various high-temperature filter devices indicate a potentially positive impact for high gradient magnetic filtration on boiling water reactor radiation level reduction. Test results on in-plant water composition and impurity crystallography are presented for several typical boiling water reactors (BWRs) on plant streams where high-temperature filtration may be particularly beneficial. An experimental model on the removal of red iron oxide (hematite) from simulated reactor water with a high gradient magnetic filter is presented, as well as the scale-up parameters used to predict the filtration efficiency on various high temperature, in-plant streams. Numerical examples are given to illustrate the crud removal potential of high gradient magnetic filters installed at alternative stream locations under typical, steady-state, plant operating conditions

  10. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    International Nuclear Information System (INIS)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment

  11. Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Dixon, D.; Lord, D.

    1978-01-01

    The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

  12. Analysis of a main steam isolation valve closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main stream isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4

  13. Theoretical investigation of flow regime for boiling water two-phase flow in horizontal rectangular narrow channels

    International Nuclear Information System (INIS)

    Zhang Chunwei; Qiu Suizheng; Yan Mingyu; Wang Bulei; Nie Changhua

    2005-01-01

    The flow regime transition criteria for the boiling water two-phase flow in horizontal rectangular narrow channels (1 x 20 mm, 2 x 20 mm) were theoretically explored. The discernible flow patterns were bubble, intermittent slug, churn, annular and steam-water separation flow. By using two-fluid model, equations of conservation of momentum were established for the two-phase flow. New flow-regime criteria were obtained and agreed well with the experiment data. (authors)

  14. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  15. Blown by the wind. Replacing nuclear power in German electricity generation

    International Nuclear Information System (INIS)

    Lechtenböhmer, Stefan; Samadi, Sascha

    2013-01-01

    Only three days after the beginning of the nuclear catastrophe in Fukushima, Japan, on 11 March 2011, the German government ordered 8 of the country's 17 existing nuclear power plants (NPPs) to stop operating within a few days. In summer 2011 the government put forward a law – passed in parliament by a large majority – that calls for a complete nuclear phase-out by the end of 2022. These government actions were in contrast to its initial plans, laid out in fall 2010, to expand the lifetimes of the country's NPPs. The immediate closure of 8 NPPs and the plans for a complete nuclear phase-out within little more than a decade, raised concerns about Germany's ability to secure a stable supply of electricity. Some observers feared power supply shortages, increasing CO 2 -emissions and a need for Germany to become a net importer of electricity. Now – a little more than a year after the phase-out law entered into force – this paper examines these concerns using (a) recent statistical data on electricity production and demand in the first 15 months after the German government's immediate reaction to the Fukushima accident and (b) reviews the most recent projections and scenarios by different stakeholders on how the German electricity system may develop until 2025, when NPPs will no longer be in operation. The paper finds that Germany has a realistic chance of fully replacing nuclear power with additional renewable electricity generation on an annual basis by 2025 or earlier, provided that several related challenges, e.g. expansion of the grids and provision of balancing power, can be solved successfully. Already in 2012 additional electricity generation from renewable energy sources in combination with a reduced domestic demand for electricity will likely fully compensate for the reduced power generation from the NPPs shut down in March 2011. If current political targets will be realised, Germany neither has to become a net electricity importer, nor will be unable

  16. Childhood cancer in the surroundings of German nuclear power plants: report of an ongoing epidemiological study

    International Nuclear Information System (INIS)

    Schulze-Rath, R.; Kaatsch, P.; Schmiedel, S.; Spix, C.; Blettner, M.

    2006-01-01

    Epidemiological studies could not show an association between ionising radiation of nuclear power stations in routine operation and the incidence of childhood cancer, yet. The following report presents a case control study conducted by the German Childhood Cancer Registry since autumn 2003. All children in the study region, who were diagnosed with cancer between 1980 and 2003 at an age below five are included. In the first part of the study we investigate whether children with cancer (cases) lived closer to the respective nuclear power stations compared to random children without cancer (controls). In the second part, for a subgroup of cases and controls we conduct computer assisted telephone interviews regarding confounders possibly associated with the exposure of ionising radiation and childhood cancer. Results are expected by the end of 2006. (orig.)

  17. The German simulator center for the training of nuclear power plant operators

    International Nuclear Information System (INIS)

    Hoffmann, E.

    1996-01-01

    Simulator training for nuclear power plant operators in Germany is conducted in The Simulator Center in Essen. The companies operating The Center are KSG/GfS. KSG provides simulators, GfS performs the training. The German Simulator Center is equipped with five simulators in training, nine simulators are under construction and will be ready for training until the beginning of 1997. This institution serves 22 nuclear power plants units in Germany, Switzerland (NPP Goesgen-Daeniken) and the Netherlands (NPP Borssele) and trains 1,800 persons every year. As a common enterprise the company is owned by 12 utilities, which leads to the necessity to prepare common rules and guidelines for simulator specification, training of instructors, assessment of trainees, training material and preparation and methodical running of simulator courses

  18. Understanding the determinants of electricity prices and the impact of the German Nuclear Moratorium in 2011

    International Nuclear Information System (INIS)

    Thoenes, Stefan

    2011-01-01

    This paper shows how the effect of fuel prices varies with the level of electricity demand. It analyzes the relationship between daily prices of electricity, natural gas and carbon emission allowances with a vector error correction model and a semiparametric varying smooth coefficient model. The results indicate that the electricity price adapts to fuel price changes in a long-term cointegration relationship. Different electricity generation technologies have distinct fuel price dependencies, which allows estimating the structure of the power plant portfolio by exploiting market prices. The semiparametric model indicates a technology switch from coal to gas at roughly 85% of maximum demand. It is used to analyze the market impact of the nuclear moratorium by the German Government in March 2011. Futures prices show that the market efficiently accounts for the suspended capacity and expects that several nuclear plants will not be switched on after the moratorium.

  19. Plant life extensions for German nuclear power plants? Controversial discussion profit taking of nuclear power plant operators

    International Nuclear Information System (INIS)

    Matthes, Felix C.

    2009-10-01

    The discussion on the plant life extensions for German nuclear power plants beyond the residual quantity of electricity particularly focus on three aspects: Effects for the emission of carbon dioxide as a greenhouse gas; Development of the electricity price for which a reduction or attenuation is postulated due to a plant life extension; Skimming of additional profits at operating companies and their use in the safeguarding of the future (development of renewable energies, support of energy efficiency, promotion of the research, consolidation of the public budget, and so on). Under this aspect, the author of the contribution under consideration reports on the profit taking of nuclear power plant operators. The presented analysis considers three aspects: (a) Specification of the quantity structures for the investigated model of plant life extension; (b) The decisive parameter is the revenue situation and thus the price development for electricity at wholesale markets; (c) Determination and evaluation of the course in time of the profit taking.

  20. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  1. Pakistan's rise to nuclear power and the contribution of German companies

    Energy Technology Data Exchange (ETDEWEB)

    Ricke, Klaus-Peter

    2013-09-01

    The amendment of the Foreign Trade and Payments Act (Aussenwirtschaftsgesetz) has prompted the preparation of this paper because of concerns over potential setbacks in advances achieved over the past twenty years in regulating German exports to non-EU countries and shipments to member states of the EU and the watering down of export restrictions to correspond to the low standards in place at the EU level (with the objective of streamlining the Foreign Trade and Payments Act and nullifying special German requirements which place German exporters at a disadvantage compared with their European competitors, according to a spokesperson of the German Federal Ministry of Economics and Technology). This would send the wrong signal on combating proliferation. From the 1970s to 1990s the Federal Republic of Germany played an extremely negative role because it opened the doors wide to the proliferation of weapons of mass destruction through lax legislation and even more slipshod enforcement. Alarmed by several scandals, in recent years the German government has taken the lead regarding this issue and it would be appropriate for it to continue to fulfill this role. The attitude of the German governments in the 1970s to 1990s as well as the attitude of the key government authorities responsible for controlling exports becomes clear through the example of the cases of illegal exports of nuclear technology to Pakistan. Until the early 1990s the legal situation was such that some acts of proliferation could not even be prosecuted because the licensing regime included only goods listed in Part 1 of the Export Control List (Ausfuhrliste Teil I). The relevant authorities - the Bundesamt fuer Wirtschaft (Federal Office of Economics, BAW) and customs authorities - were neither in terms of personnel nor due to the deficiency of their computer equipment - in a position to carry out their legally prescribed duties. The maximum sentence for even the most serious proliferation violations

  2. Pakistan's rise to nuclear power and the contribution of German companies

    International Nuclear Information System (INIS)

    Ricke, Klaus-Peter

    2013-01-01

    The amendment of the Foreign Trade and Payments Act (Aussenwirtschaftsgesetz) has prompted the preparation of this paper because of concerns over potential setbacks in advances achieved over the past twenty years in regulating German exports to non-EU countries and shipments to member states of the EU and the watering down of export restrictions to correspond to the low standards in place at the EU level (with the objective of streamlining the Foreign Trade and Payments Act and nullifying special German requirements which place German exporters at a disadvantage compared with their European competitors, according to a spokesperson of the German Federal Ministry of Economics and Technology). This would send the wrong signal on combating proliferation. From the 1970s to 1990s the Federal Republic of Germany played an extremely negative role because it opened the doors wide to the proliferation of weapons of mass destruction through lax legislation and even more slipshod enforcement. Alarmed by several scandals, in recent years the German government has taken the lead regarding this issue and it would be appropriate for it to continue to fulfill this role. The attitude of the German governments in the 1970s to 1990s as well as the attitude of the key government authorities responsible for controlling exports becomes clear through the example of the cases of illegal exports of nuclear technology to Pakistan. Until the early 1990s the legal situation was such that some acts of proliferation could not even be prosecuted because the licensing regime included only goods listed in Part 1 of the Export Control List (Ausfuhrliste Teil I). The relevant authorities - the Bundesamt fuer Wirtschaft (Federal Office of Economics, BAW) and customs authorities - were neither in terms of personnel nor due to the deficiency of their computer equipment - in a position to carry out their legally prescribed duties. The maximum sentence for even the most serious proliferation violations

  3. Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

    International Nuclear Information System (INIS)

    Schulz, W.

    2000-01-01

    concept - to calculate the control rod's working life both in the control position and the shut-down position - will automatically lead to an optimization of the control rod strategy. Control rod optimisation is demonstrated by accumulating the total amount of control rods required in a medium-sized BWR up to the total reactor holding period. At least 60% of the first core inventory - for this control rod type an existing EMPIRICAL MODEL is already available - may be used up to the total operating period without any safety loss. Looking to the present disposal situation this concept represents a practical way to reduce all high level waste. In addition benefit of utilizing this concept is that it minimizes tritium emission. Control-rods utilized within Boiling Water Reactors (BWR) are designed for the purpose to control and shape the neutron flux profile in the reactor, to adjust the range of regulation referring to the weight rate of the reactor coolant and thirdly by- shutting down the reactor at any time and under any conditions with regard to nuclear aspects, mechanical integrity and control rod history. The designation control- or shut down rod characterize the particular field of activity for a given control rod. The focal point of my work had shown to be a calculation of the nuclear working life of any control rod design as well as an optimisation method with reference to the holding period for a given control rod inventory as a result of measuring data and a theoretical analysis describing the parameters in a general validity form. (author)

  4. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der.

    1989-01-01

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  5. Study of the radiotoxicity of actinides recycling in boiling water reactors fuel

    International Nuclear Information System (INIS)

    Francois, J.L.; Guzman, J.R.; Martin-del-Campo, C.

    2009-01-01

    In this paper the production and destruction, as well as the radiotoxicity of plutonium and minor actinides (MA) obtained from the multi-recycling of boiling water reactors (BWR) fuel are analyzed. A BWR MOX fuel assembly, with uranium (from enrichment tails), plutonium and minor actinides is designed and studied using the HELIOS code. The actinides mass and the radiotoxicity of the spent fuel are compared with those of the once-through or direct cycle. Other type of fuel assembly is also analyzed: an assembly with enriched uranium and minor actinides; without plutonium. For this study, the fuel remains in the reactor for four cycles, where each cycle is 18 months length, with a discharge burnup of 48 MWd/kg. After this time, the fuel is placed in the spent fuel pool to be cooled during 5 years. Afterwards, the fuel is recycled for the next fuel cycle; 2 years are considered for recycle and fuel fabrication. Two recycles are taken into account in this study. Regarding radiotoxicity, results show that in the period from the spent fuel discharge until 1000 years, the highest reduction in the radiotoxicity related to the direct cycle is obtained with a fuel composed of MA and enriched uranium. However, in the period after few thousands of years, the lowest radiotoxicity is obtained using the fuel with plutonium and MA. The reduction in the radiotoxicity of the spent fuel after one or two recycling in a BWR is however very small for the studied MOX assemblies, reaching a maximum reduction factor of 2.

  6. Corrosion products, activity transport and deposition in boiling water reactor recirculation systems

    International Nuclear Information System (INIS)

    Alder, H.P.; Buckley, D.; Grauer, R.; Wiedemann, K.H.

    1989-09-01

    The deposition of activated corrosion products in the recirculation loops of Boiling Water Reactors produces increased radiation levels which lead to a corresponding increase in personnel radiation dose during shut down and maintenance. The major part of this dose rate is due to cobalt-60. The following areas are discussed in detail: - the origins of the corrosion products and of cobalt-59 in the reactor feedwaters, - the consolidation of the cobalt in the fuel pin deposits (activation), - the release and transport of cobalt-60, - the build-up of cobalt-60 in the corrosion products in the recirculation loops. Existing models of the build-up of circuit radioactivity are discussed and the operating experiences from selected reactors are summarised. Corrosion chemistry aspects of the cobalt build-up in the primary circuit have already been studied on a broad basis and are continuing to be researched in a number of centers. The crystal chemistry of chromium-nickel steel corrosion products poses a number of yet unanswered questions. There are major loopholes associated with the understanding of activation processes of cobalt deposited on the fuel pins and in the mass transfer of cobalt-60. For these processes, the most important influence stems from factors associated with colloid chemistry. Accumulation of data from different BWRs contributes little to the understanding of the activity build-up. However, there are examples that the problem of activity build-up can be kept under control. Although many details for a quantitative understanding are still missing, the most important correlations are visible. The activity build-up in the BWR recirculation systems cannot be kept low by a single measure. Rather a whole series of measures is necessary, which influences not only cobalt-60 deposition but also plant and operation costs. (author) 26 figs., 13 tabs., 90 refs

  7. Investigation of noble metal deposition behaviour in boiling water reactors. The NORA project

    International Nuclear Information System (INIS)

    Ritter, Stefan; Karastoyanov, Vasil; Abolhassani-Dadras, Sousan; Guenther-Leopold, Ines; Kivel, Niko

    2010-01-01

    NobleChem trademark is a technology developed by General Electric to reduce stress corrosion cracking (SCC) in reactor internals and recirculation pipes of boiling water reactors (BWRs) while preventing the negative side effects of classic hydrogen water chemistry. Noble metals (Pt, Rh) acting as electrocatalysts for the recombination of O 2 and H 2 O 2 with H 2 to H 2 O and thus reducing the corrosion potential more efficiently are injected into the feedwater during reactor shutdown (classic method) or on-line during power operation. They are claimed to deposit as very fine metallic particles on all water-wetted surfaces, including the most critical regions inside existing cracks, and to stay electrocatalytic over long periods of time. The effectiveness of this technology in plants still remains to be demonstrated. Based on highly credible laboratory experiments down to the sub-μg . kg -1 Pt concentration range, SCC mitigation may be expected, provided that a stoichiometric excess of H 2 and a sufficient surface coverage with very fine Pt particles exist simultaneously at the critical locations [1]. Very little is known about the deposition and (re-)distribution behaviour of the Pt in the reactor. For the validation of this technique the research project NORA (noble metal deposition behaviour in BWRs) has been started at the Paul Scherrer Institute (PSI) with two main objectives: (i) to gain phenomenological insights and a better basic understanding of the Pt distribution and deposition behaviour in BWRs; (ii) to develop and qualify a non-destructive technique to characterise the size and distribution of the Pt particles and the local concentration of Pt on reactor components. This paper presents the objectives of the project, the planned work and a brief description of the status of the project. (orig.)

  8. Investigation of the noble metal deposition behaviour in boiling water reactors - the NORA project

    International Nuclear Information System (INIS)

    Ritter, S.; Karastoyanov, V.; Abolhassani-Dadras, S.; Guenther-Leopold, I.; Kivel, N.

    2010-01-01

    NobleChem™ is a technology developed by General Electric to reduce stress corrosion cracking (SCC) in reactor internals and recirculation pipes of boiling water reactors (BWRs) while preventing the negative side effects of classical hydrogen water chemistry. Noble metals (Pt, Rh) acting as electrocatalysts for the recombination of O 2 and H 2 O 2 with H 2 to H 2 O and thus reducing the corrosion potential more efficiently are injected into the feed water during reactor shut-down (classical method) or on-line during power operation. They are claimed to deposit as very fine metallic particles on all water-wetted surfaces including the most critical regions inside existing cracks and to stay electrocatalytic over long periods of time. The effectiveness of this technology in plants remains still to be demonstrated. Based on highly credible laboratory experiments down to the sub-ppb Pt concentration range, SCC mitigation may be expected, provided that a stoichiometric excess of H 2 and a sufficient surface coverage with very fine Pt particles exist simultaneously at the critical locations. Very little is known about the deposition and (re-)distribution behaviour of the Pt in the reactor. For the validation of this technique the research project NORA (noble metal deposition behaviour in BWRs) has been started at PSI with two main objectives: (i) to gain phenomenological insights and a better basic understanding of the Pt distribution and deposition behaviour in BWRs; (ii) to develop and qualify a non-destructive technique to characterise the size and distribution of the Pt particles and its local concentration on reactor components. This paper presents the objectives of the project, the planned work and a brief description of the status of the project. (author)

  9. Simulation of heat and mass transfer in boiling water with the Melodif code

    International Nuclear Information System (INIS)

    Freydier, P.; Chen, O.; Olive, J.; Simonin, O.

    1991-04-01

    The Melodif code is developed at Electricite de France, Research and Development Division. It is an eulerian two dimensional code for the simulation of turbulent two phase flows (a three dimensional code derived from Melodif, ASTRID, is currently being prepared). Melodif is based on the two fluid model, solving the equations of conservation for mass, momentum and energy, for both phases. In such a two fluid model, the description of interfacial transfers between phases is a crucial issue. The model used applies to a dominant continuous phase, and a dispersed phase. A good description of interfacial momentum transfer exists in the standard MELODIF code: the drag force, the apparent mass force... are taken into account. An important factor for interfacial transfers is the interfacial area per volume unit. With the assumption of spherical gas bubbles, an equation has been written for this variable. In the present wok, a model has been tested for interfacial heat and mass transfer in the case of boiling water: it is assumed that mass transfer is controlled by heat transfer through the latent massic energy taken in the phase that vaporizes (or condenses). This heat and mass transfer model has been tested in various configurations: - a cylinder with water flowing inside, is being heated. Boiling takes place near the wall, while bubbles migrating to the core of the flow recondense. This roughly simulates a sub-cooled boiling phenomenon. - a box containing liquid water is depressurized. Boiling takes place in the whole volume of the fluid. The Melodif code can simulate this configuration due to the implicitation of the relation between interphase mass transfer and the pressure variable

  10. Implementation of automated, on-line fatigue monitoring in a boiling water reactor

    International Nuclear Information System (INIS)

    Sakai, Takeshi; Tokunaga, Katsumi; Stevens, G.L.; Ranganath, S.

    1993-01-01

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to a Japanese operating boiling water reactor (BWR), Tsuruga Unit 1, is described. The system uses the influence function approach and rainflow cycle counting methodology, operates on a workstation computer, and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant-unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computes the fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification number-sign 501. Fatigue values are saved automatically on files at times defined by the user for use at a later time. Of particular note, this paper describes some of the details involved with implementing such a system from the utility perspective. Utility installation details, as well as why such a system was chosen for implementation are presented. Fatigue results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. Although the system is specifically set up to address fatigue duty for the feedwater nozzle location, a generic shell structure was implemented so that any other components could be added at a future time without software modifications. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension

  11. A boiling-water reactor concept for low radiation exposure based on operating experience

    International Nuclear Information System (INIS)

    Koine, Y.; Uchida, S.; Izumiya, M.; Miki, M.

    1983-01-01

    A review of boiling-water reactor (BWR) operating experience indicates the significant role of water chemistry in determining the radiation dose rate contributing to occupational exposure. The major contributor among the radioactive species involved is identified as 60 Co, produced by neutron activation of 59 Co originating from structural materials. Iron crud, a fine solid form of corrosion product in the reactor water, is also shown to enhance the radiation dose rate. A theoretical study, supported by the operating experience and an extensive confirmatory test, led to the computerized analytical model called DR CRUD which is capable of predicting long-term radiation dose buildup. It accounts for the mechanism of radiation buildup through corrosion products such as irons, cobalts and other radioactive elements; their generation, transport, activation, interaction and deposition in the reactor coolant system are simulated. A scoping analysis, using this model as a tool, establishes the base line of the BWR concept for low occupational exposure. The base line consists of a set of target values for an annual exposure of 200 man.rem in an 1100 MW(e) BWR unit. They are the parameters that will be built into the design such as iron and cobalt inputs to the reactor water, and the capability of the reactor and the condensate purification system. Applicable means of technology are identified to meet the targets, ranging from improved water chemistry to the purification technique, optimized material selection and the recommended operational procedure. Extensive test programmes provide specifications of these means for use in BWRs. Combinations of their application are reviewed to define the concept of reduced exposure. Analytical study verifies the effectiveness of the proposed BWR concept in achieving a low radiation dose rate; occupational exposure is reduced to 200 man.rem/a. (author)

  12. Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

    1995-09-01

    Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

  13. Parametric study of recriticality in a boiling water reactor severe accident

    International Nuclear Information System (INIS)

    Shamoun, B.I.; Witt, R.J.

    1994-01-01

    Recriticality is possible in a severe accident if unborated or low boron concentration water is added to a damaged core after control rod melting but before fuel melting. Recriticality in a severe accident in a boiling water reactor was parametrically investigated using the TWODANT code. Eigenvalue calculations for a unit central fuel cell with reflective boundary conditions were performed by solving the two-dimensional multigroup steady-state Boltzman transport equation using TWODANT. Two sets of calculations were performed in this work. The first set of calculations was carried out under three types of normal operating conditions to provide reference values for the accident calculations: (a) cold rodded condition, (b) cold unrodded condition, and (c) hot full-power condition. The eigenvalues at these conditions were found to be 1.055, 1.208, and 1.098, respectively. The second set of calculations was carried out after the melting of the control element and during the reflood phase, under the following reflood conditions: (a) reflood with unborated water and (b) reflood with borated water. For the reflood case with unborated water, five values of void fractions were considered (100, 60, 40, 20, and 0%). Decreasing void fractions represent greater refill levels during the reflood process. The system pressure was taken to be 7 MPa, while the moderator temperature was set to 560 K. Plotting the eigenvalue compared with the fraction of control materials lost indicates recriticality is only possible if nearly 100% of the control material is lost from the core. Eigenvalue calculations were repeated for short- and long-term recovery conditions of the reflood phase corresponding to maximum moderator density at 4 MPa pressure and 525 K moderator temperature and for 1 MPa pressure and 325 K moderator temperature, respectively. Recriticality was again observed to be a concern only after losing 95% ore more of control materials from the unit cell

  14. Involving private persons and corporations in the planning process according to paragraph 9b German nuclear law - practicability and limits

    International Nuclear Information System (INIS)

    Altmann, J.; Roesel, H.

    1978-01-01

    The 4th Amendment to the German Nuclear Law has procured the legal base for the disposal of radioactive waste in the German Federal Republic. The Nuclear Law states that the Physikalisch-Technische Bundesanstalt (PTB) in Braunschweig is responsible for the disposal of radioactive waste. The PTB can be assisted in its responsibilities by persons and institutions of private character. Involving private persons and corporations in the licensing process figures a legal problem, and there has been a seminar in the PTB discussing such problems. The present contribution gives a brief summary of the opinions stated by the various legal experts taking part in this seminar. (orig./HP) [de

  15. Technology, safety and costs of decommissioning a refernce boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  16. International nuclear safety experts complete IAEA peer review of German regulatory system

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: An international expert team has today completed a two-week IAEA review of Germany's nuclear regulatory system. The team identified good practices within the system and gave advice on some areas for further improvement. The IAEA has conveyed the initial findings to German authorities but the final report will be submitted within two months. At the request of the Government of the Federal Republic of Germany, the International Atomic Energy Agency (IAEA) assembled a team of 14 experts to conduct an Integrated Regulatory Review Service (IRRS) mission. This is a peer review based on IAEA Standards. It is not an inspection, nor an audit. The scope of the mission was limited to the safety regulation of nuclear power plants. Experts from Canada, the Czech Republic, Finland, France, Japan, the Netherlands, Republic of Korea, Spain, Switzerland, the UK, the US and from the IAEA took part in the mission, which was conducted from 7 to 19 September in Bonn, Stuttgart and Berlin. The main basis for the review was a well-prepared self-assessment made by the Federal Ministry of Environment, Nature Conservation and Nuclear Safety (BMU) and the Ministry of Environment of the federal state of Baden-Wuerttemberg (UM BW). 'The team members were impressed by the extensive preparation and dedication of the staff both at BMU and UM BW to excellence in nuclear safety,' said Mike Weightman, IRRS Team Leader and Chief Inspector of the UK nuclear regulatory body, the Nuclear Directorate of the Health and Safety Executive. 'We hope the IRRS mission will facilitate further improvements in the safety regulation of nuclear power in Germany and throughout the world.' 'Germany's invitation to undergo such a detailed review is a clear demonstration of its openness and commitment to continuously improve nuclear safety regulation,' said Philippe Jamet, Director of the IAEA's Nuclear Installation Safety Division. Among the particular strengths of BMU and UM BW associated with their

  17. German nuclear codes revised: comparison with approaches used in other countries

    International Nuclear Information System (INIS)

    Raetzke, C.; Micklinghoff, M.

    2005-01-01

    The article deals with the plan of the German Federal Ministry for the Environment (BMU) to revise the German set of nuclear codes, and draws a comparison with approaches pursued in other countries in formulating and implementing new requirements imposed upon existing plants. A striking feature of the BMU project is the intention to have the codes reflect the state of the art in an entirely abstract way irrespective of existing plants. This implies new requirements imposed on plant design, among other things. However, the state authorities, which establish the licensing conditions for individual plants in concrete terms, will not be able to apply these new codes for legal reasons (protection of vested rights) to the extent in which they incorporate changes in safety philosophy. Also the procedure adopted has raised considerable concern. The processing time of two years is inordinately short, and participation of the public and of industry does not go beyond the strictly formal framework of general public participation. In the light of this absence of quality assurance, it would be surprising if this new set of codes did not suffer from considerable deficits in its contents. Other countries show that the BMU is embarking on an isolated approach in every respect. Elsewhere, backfitting requirements are developed carefully and over long periods of time; they are discussed in detail with the operators; costs and benefits are weighted, and the consequences are evaluated. These elements are in common to procedures in all countries, irrespective of very different steps in detail. (orig.)

  18. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Annuli (Part I)

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Hernborg, G.

    1962-12-01

    The present report deals with measurements of burnout conditions for flow of boiling water in an annulus with an inner diameter of 9.92 mm, an outer diameter of 17 - 42 mm and a heated length of 608 mm. Data were obtained in respect of external heating only, internal heating only and dual uniform and non-uniform heating. The following ranges of variables were studied and 978 burnout measurements were obtained. Pressure 8.5 2 ; Inlet subcooling 60 sub i 2 ; Outer surface heat flux 0 o 2 ; Mass velocity 71 2 /sec; The results are presented in diagrams where the burnout steam qualities, x BO , were plotted against the pressure with the surface heat fluxes as parameters. The data have been correlated by curves. The scatter of the data around the curves is less than ± 5 per cent. In the case of equal heat fluxes on both walls of the annulus, burnout always occurred on the inner wall, and the data compared rather well with round duct data. When the annulus was heated internally only, the data showed very low burnout values in comparison with the results for dual heating and round ducts. This disagreement was explained by considering the climbing film flow model and by the fact that only a fraction of the channel perimeter was heated. For external heating the data are somewhat lower than corresponding round duct data, but rather high in comparison with internal heating. The climbing film flow model was also used to interpret this observation. For dual non-uniform heating it was found that the outer surface may be overloaded from 30 to 70 per cent compared with the inner surface without reducing the margin of safety in respect of burnout for the annulus. It was further observed that when the heat flux fox the wall on which burnout occurs is increased, the burnout steam quality for the channel decreases. If, however, the heat flux for the opposite wall is increased, the burnout steam quality also increases. It was also observed that the highest burnout values are obtained

  19. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Annuli (Part I)

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Hernborg, G

    1962-12-15

    The present report deals with measurements of burnout conditions for flow of boiling water in an annulus with an inner diameter of 9.92 mm, an outer diameter of 17 - 42 mm and a heated length of 608 mm. Data were obtained in respect of external heating only, internal heating only and dual uniform and non-uniform heating. The following ranges of variables were studied and 978 burnout measurements were obtained. Pressure 8.5 < 37.5 kg/cm{sup 2}; Inlet subcooling 60 < {delta}t{sub sub} < 205 deg C; Steam quality 0.1 < x < 0.91; Inner surface heat flux 0 < (q/A){sub i} < 303 W/cm{sup 2}; Outer surface heat flux 0 < (q/A){sub o} < 374 W/cm{sup 2}; Mass velocity 71 < m/F < 961 kg/m{sup 2}/sec; The results are presented in diagrams where the burnout steam qualities, x{sub BO}, were plotted against the pressure with the surface heat fluxes as parameters. The data have been correlated by curves. The scatter of the data around the curves is less than {+-} 5 per cent. In the case of equal heat fluxes on both walls of the annulus, burnout always occurred on the inner wall, and the data compared rather well with round duct data. When the annulus was heated internally only, the data showed very low burnout values in comparison with the results for dual heating and round ducts. This disagreement was explained by considering the climbing film flow model and by the fact that only a fraction of the channel perimeter was heated. For external heating the data are somewhat lower than corresponding round duct data, but rather high in comparison with internal heating. The climbing film flow model was also used to interpret this observation. For dual non-uniform heating it was found that the outer surface may be overloaded from 30 to 70 per cent compared with the inner surface without reducing the margin of safety in respect of burnout for the annulus. It was further observed that when the heat flux fox the wall on which burnout occurs is increased, the burnout steam quality for the

  20. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

  1. A non-linear reduced order methodology applicable to boiling water reactor stability analysis

    International Nuclear Information System (INIS)

    Prill, Dennis Paul

    2013-01-01

    Thermal-hydraulic coupling between power, flow rate and density, intensified by neutronics feedback are the main drivers of boiling water reactor (BWR) stability behavior. High-power low-flow conditions in connection with unfavorable power distributions can lead the BWR system into unstable regions where power oscillations can be triggered. This important threat to operational safety requires careful analysis for proper understanding. Analyzing an exhaustive parameter space of the non-linear BWR system becomes feasible with methodologies based on reduced order models (ROMs), saving computational cost and improving the physical understanding. Presently within reactor dynamics, no general and automatic prediction of high-dimensional ROMs based on detailed BWR models are available. In this thesis a systematic self-contained model order reduction (MOR) technique is derived which is applicable for several classes of dynamical problems, and in particular to BWRs of any degree of details. Expert knowledge can be given by operational, experimental or numerical transient data and is transfered into an optimal basis function representation. The methodology is mostly automated and provides the framework for the reduction of various different systems of any level of complexity. Only little effort is necessary to attain a reduced version within this self-written code which is based on coupling of sophisticated commercial software. The methodology reduces a complex system in a grid-free manner to a small system able to capture even non-linear dynamics. It is based on an optimal choice of basis functions given by the so-called proper orthogonal decomposition (POD). Required steps to achieve reliable and numerical stable ROM are given by a distinct calibration road-map. In validation and verification steps, a wide spectrum of representative test examples is systematically studied regarding a later BWR application. The first example is non-linear and has a dispersive character

  2. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Science.gov (United States)

    Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

    2018-01-01

    Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

  3. Variation of the Effectiveness of Hydrogen Water Chemistry in a Boiling Water Reactor during Startup Operations

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya

    2012-09-01

    For mitigating intergranular stress corrosion cracking (IGSCC) in an operating boiling water reactor (BWR), the technology of hydrogen water chemistry (HWC) aiming at coolant chemistry improvement has been adopted worldwide. However, the hydrogen injection system employed in this technology was designed to operate only at power levels greater than 30% of the rated power or at coolant temperatures of greater than 450 deg. F. This system is usually in an idle and standby mode during a startup operation. The coolant in a BWR during a cold shutdown normally contains a relatively high level of dissolved oxygen from intrusion of atmospheric air. Accordingly, the structural materials in the primary coolant circuit (PCC) of a BWR could be exposed to a strongly oxidizing environment for a short period of time during a subsequent startup operation. At some plants, the feasibility of hydrogen water chemistry during startup operations has been studied, and its effectiveness on suppressing SCC initiation was evaluated. It is technically difficult to directly procure water chemistry data at various locations of an operating reactor. Accordingly, the impact of startup operation on water chemistry in the PCC of a BWR operating under normal water chemistry (NWC) or HWC can only be theoretically evaluated through computer modelling. In this study, a well-developed computer code DEMACE was used to investigate the variations in redox species concentration and in electrochemical corrosion potential (ECP) of components in the PCC of a domestic BWR during startup operations in the presence of HWC. Simulations were carried out for [H2] FW s ranging from 0.0 to 2.0 parts per million (ppm) and for power levels ranging from 2.5% to 11.3% during startup operations. Our analyses indicated that for power levels with steam generation in the core, a higher power level would tend to promote a more oxidizing coolant environment for the structural components and therefore lead to less HWC

  4. German energy turnaround and Poland's start in nuclear power? A survey

    International Nuclear Information System (INIS)

    Knopp, Lothar; Gorski, Marek

    2015-01-01

    The book includes contributions from two meetings: 1. Meeting in Berlin, October 8,2014: Energy transition in Poland - historical background, development and actual situation; legal boundary conditions of the entry into nuclear power in Poland; ecologic and economic causes motivation for the atomic energy entry in Poland; promotion systems for renewable energy in Poland, current status and political background; construction of a nuclear power plant in Northern Poland site specific impacts and social assessment. Meeting in Stettin on November 5, 2014: Energy turnaround in Germany -German nuclear phaseout under consideration of EU legislation; ecologic and economic motivation for the energy turnaround in Germany; energy turnaround in Germany - new legal boundary conditions, especially the amendment to tue renewable energy law; contribution of the industry to the energy turnaround - taking the example waste management and waste incineration plants; fracking as a factor of the energy turnaround? - legal boundary conditions and ecologic risks; ecological, ethical and sociopolitical aspects of the energy turnaround in Germany and Poland.

  5. Nuclear Experts Complete IAEA Follow-up Review of German Regulatory System

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: Nuclear safety experts concluded a seven-day mission to review the German Regulatory System, conducted from 4-10 September in Bonn, Stuttgart and Berlin. At the request of the Government of the Federal Republic of Germany, the International Atomic Energy Agency assembled a peer-review team of six high-level regulatory experts from six nations (Finland, France, the Netherlands, Switzerland, the UK, the US and three IAEA senior staff members) to conduct a follow-up assessment of an Integrated Regulatory Review Service (IRRS) mission conducted in 2008. This follow-up IRRS mission examined the progress in acting upon the recommendations and suggestions made during the 2008 IRRS mission and reviewed the areas of significant regulatory changes since that review at both the Federal Ministry of Environment, Nature Conservation and Nuclear Safety (BMU) and the Ministry of Environment of the federal state of Baden-Wurttemberg (UM BW). The first mission reviewed Germany's regulatory framework against IAEA Safety Standards and fostered the exchange of information and experience on safety regulation. This is a peer review based on IAEA Standards. It is not an inspection, nor an audit. The scope of the mission was limited to the safety regulation of nuclear power plants. IRRS team leader, Mr. McCree, of the US Nuclear Safety Commission (USNRC), said, ''This was an important IRRS mission, particularly given the recent Fukushima Daiichi Nuclear Power Plant accident and the related insights which underscore the importance of having an independent, credible nuclear safety regulator.'' ''The IRRS team identified several strengths of the German nuclear safety regulators, including the prompt and coordinated incident response activities of BMU and UM BW to the Fukushima accident. Some suggestions were also made to further strengthen nuclear safety regulations concerning the future work of BMU,'' he said. The review team found that important progress has been made toward

  6. The German energy policy as a consequence of Fukushima. The scientific discussion between nuclear phase-out and economic growth

    International Nuclear Information System (INIS)

    Radtke, Joerg

    2013-01-01

    The book on the German energy policy as a consequence includes the following contributions: The German energy turnaround - scientific contributions. The energy turnaround in Germany - issue of interdisciplinary science. The transformation of the energy systems as social and technical challenge, - on the need of integrating energy research. Transformations and transformation blockades in the German energy system. The German energy turnaround in the context of international best practice. Energy turnaround also in Japan? - The chances of a nuclear phase-out. Possibilities and limits of public participation for the realization of an energy turnaround. Public energy in Germany - a model for participation? A plea for a comprehensive analysis of the energy turnaround in relation to the omnipresent crisis. Challenges and development in the German energy industry - consequences of the increasing percentage of renewable energies on the costs and the security of supply. Research funding and innovation promotion in the area of selected renewable energies. The economic chances of an energy turnaround. The need of appropriate monetary boundary conditions for the energy turnaround and the possibilities of an organization. The human factor in the context of the energy turnaround - environmental-psychological research approaches. The legal contribution to the energy turnaround. Vulnerability and resilience of energy systems. Geography of renewable energies -spatial constraints of a sustainable energy system. Critics and alternatives: The German energy turnaround that is no turnaround.

  7. Corrosion product deposition on fuel element surfaces of a boiling water reactor

    International Nuclear Information System (INIS)

    Orlov, A.

    2011-01-01

    Over the last decade the problem of corrosion products deposition on light water reactor fuel elements has been extensively investigated in relation to the possibility of failures caused by them. The goal of the present study is to understand in a quantitative way the formation of such kind of deposits and to analytically understand the mechanism of formation and deposition with help of the quasi-steady state concentrations of a number of 3d metals in reactor water. Recent investigations on the complex corrosion product deposits on a Boiling Water Reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The buildup of magnetic corrosion product deposits (crud) on the fuel cladding of the BWR, Kernkraftwerk Leibstadt (KKL) Switzerland has hampered the Eddy-current based measurements of ZrO 2 layer thickness. The magnetic behavior of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main components identified by 2D μXRD mapping inside the layer were hematite and spinel phases with the common formula (M x Fe y )[M (1-x) Fe (2-y) ]O 4 , where M = Zn, Ni, Mn. With μXRD it was clearly shown that the cell parameter of analyzed spinel is different from the one of the pure endmembers (ZnFe 2 O 4 , NiFe 2 O 4 and MnFe 2 O 4 ) proving the existence of

  8. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  9. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  10. A new coupling kernel for the three-dimensional simulation of a boiling water reactor core by the nodal coupling method

    International Nuclear Information System (INIS)

    Gupta, N.K.

    1981-01-01

    A new coupling kernel is developed for the three-dimensional (3-D) simulation of Boiling Water Reactors (BWR's) by the nodal coupling method. The new kernel depends not only on the properties of the node under consideration but also on the properties of its neighbouring nodes. This makes the kernel more useful in particular for fuel bundles lying in a surrounding of different nuclear characteristics, e.g. for a controlled bundle in the surrounding of uncontrolled bundles or vice-versa. The main parameter in the new kernel is a space-dependent factor obtained from the ratio of thermal-to-fast flux. The average value of the above ratio for each node is evaluated analytically. The kernel is incorporated in a 3-D BWR core simulation program MOGS. As an experimental verification of the model, the cycle-6 operations of the two units of the Tarapur Atomic Power Station (TAPS) are simulated and the result of the simulation are compared with Travelling Incore Probe (TIP) data. (orig.)

  11. Anticipated transient without scram analysis of the simplified boiling water reactor following main steam isolation valve closure with boron injection

    International Nuclear Information System (INIS)

    Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

    1996-01-01

    The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system

  12. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  13. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs

  14. Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

    International Nuclear Information System (INIS)

    Nunez C, A.; Espinosa P, G.

    2007-01-01

    Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

  15. Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

    International Nuclear Information System (INIS)

    Panajotov, D.P.; Gorburov, V.I.

    1989-01-01

    Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

  16. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1978-01-01

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

  17. German-German energy industry - on the future cooperation of enterprises in the field of nuclear energy

    International Nuclear Information System (INIS)

    Strassburg, W.

    1990-01-01

    A joint venture for a nuclear power plant in operation or under construction poses for more open questions than a joint venture for a western-standard plant still to be erected. Both solutions have in common that resistance against the peaceful uses of nuclear energy in the former GDR is groving. It is too early to make a statement concerning joint ventures regarding other nuclear facilities, e.g. final repositories. That there will be no 'nuclear export' as such after unification does not imply a lot. The changes in eastern Europe are one of the biggest challenges in history. They are to be conquered by joining forces. This also goes for nuclear energy where not only safety but also the environment is at stake. Legislation should pay appropriate heed to these problems. (orig./HSCH) [de

  18. 2008 annual nuclear technology conference: opting out of the use of nuclear power. German special approach leads into a dead end of energy policy. Conference report

    International Nuclear Information System (INIS)

    Anon.

    2008-01-01

    The President of the Deutsches Atomforum, Dr. Walter Hohlefelder, emphasized in his opening address at the 2008 Annual Nuclear Technology Conference in Hamburg that the German special approach to nuclear power utilization led straight into a dead end of energy policy. ''The outcome is foreseeable: The ambitious German goals of carbon dioxide reduction are missed, the competitiveness of the country is jeopardized, dependency on foreign energy imports rises,'' Dr. Hohlefelder stated. In view of the growing challenges in energy policy Germany had no alternative but to reassess nuclear power. The only outcome of this reappraisal could be extension of the life of nuclear power plants currently in operation. This was necessary also in order to avoid an impending gap in German electricity supply, Dr. Hohlefelder added. He invited all stakeholders to join in an open, unbiased dialog. Dr. Hohlefelder openly criticized the continued ban on research into the development of new reactors. ''A policy of this kind, a policy which bans thinking, is unacceptable in a technology-oriented, industrialized nation such as Germany.'' Nuclear power technology as a high-tech area was a unique achievement which had contributed to the prosperity of the country. The Annual Nuclear Technology Conference, which was held for the 39th time this year, is one of the biggest specialized conferences in the nuclear field with an attendance, this year, of approximately 1300 participants from more than twenty nations. (orig.)

  19. Nuclear power and the opting-out policy of the German Government. A matter of responsibility and controversy

    International Nuclear Information System (INIS)

    Grawe, J.; Voss, A.

    2000-01-01

    The article critically reviews the energy policy of the German Government and the plans for winding down nuclear power generation and instead pursue a policy for a ''green reshufflement'' of generation technologies, providing for incentives to switch to options for decentralized power generation with enhanced use of renewable energy sources. The article explains arguments against the Government's opting-out policy which were laid down in a memorandum presented to the Government by a group of five university professors, among them the authors of this article who, as experts in this domain, are aware of their responsibility as scientists and members of society and therefore explain their view of the policy issues, also referring to a recent expert opinion of the German Wissenschaftsrat (science council), which recommended to continue nuclear energy research programmes and nuclear power generation. (orig./CB) [de

  20. Statistical inquiry on the reliability of emergency diesel stations in German nuclear power plants

    International Nuclear Information System (INIS)

    1983-01-01

    This statistic inquiry is based on 692 occurrances in 40 diesel stations of 10 German nuclear power plants. Various parameters influencing the failure behaviour of diesel stations were investigated on only significant plant-specific influences and the impact of diesel station circuitry on failure behaviour were established. According to the results of this inquiry, running time, start-up number and increasing operational experience do not apparently influence the failure behaviour of diesel stations. The expected failure probability of diesel stations varies with the different nuclear power plants. Taking into account both start-up and operational failures, (with monthly inspections and running times of up to 2 h), this value is in the range of 1.6 x 10 -2 to 1.7 x 10 -3 per application. Considering failure data of all diesel stations, the failure probability (start-up and operational failures) is 8.1 x 10 -3 per application. On account of the two common-mode failures registered, a common-mode failure probability of 10 -3 was established. The inquiry also showed that non-availability of diesel stations is essentially determined by maintenance intervals. (orig.) [de

  1. Fundamentals, constitution and tasks of the German authorities for quality assurance in nuclear medicine

    International Nuclear Information System (INIS)

    Rink, T.

    2005-01-01

    Among other novelties, the national realization of the European Council Directive 97/43 Euratom demanded the formation of competent authorities for quality assurance in nuclear medicine that have become operative in the meantime. Their tasks are laid down in the revised versions of the German ''Strahlenschutzverordnung'' and ''Richtlinie Strahlenschutz in der Medizin''. They perform periodical investigations in all nuclear medicine installations to verify compliance with legal provisions on radiological protection, functioning as independent mediators between practitioners and persons legally responsible for the medical facilities on the one hand, and the supervisory board of the national administration on the other hand. They assist physicians and managers with fulfillment of the legal requirements and give advise for optimizing medical radiological procedures. The government has to be notified about the summarized results of the regular examinations at least once a year. However, severe infringements, consistent excess of the diagnostic reference levels or refusal to contribute the demanded documents by a facility undergoing an inspection, have to be reported immediately. This section describes the legal fundamentals, the constitution and the tasks of the inspecting authorities in Germany with accentuation on the course of the investigations. (orig.)

  2. A novel start-up procedure for natural-circulation boiling water reactors

    International Nuclear Information System (INIS)

    Annalisa Manera; Frank Schaefer

    2005-01-01

    Full text of publication follows: The elimination of recirculation pumps and associated systems, as proposed for natural-circulation Boiling Water Reactors (BWRs), allow a great simplification in the design of BWRs. On the other hand, it has been shown both experimentally and analytically that such a new reactor configuration makes the system susceptible to thermal-hydraulic instabilities during the start-up phase (so-called flashing-induced instabilities). Therefore, appropriate start-up procedures have to be planned to avoid instabilities in natural-circulation BWRs. Not many proposals of start-up procedures for natural-circulation BWRs are reported in literature, but all authors agree on the fact that the system should be pressurized before the transition to two-phase circulation is allowed. Nayak [1] and Jiang and coauthors [2] proposed to externally pressurize the system by injecting in the pressure vessel respectively steam produced in a separate boiler or nitrogen. Once the pressure in the reactor vessel is high enough, the reactor power can be increased to achieve two-phase natural circulation. Unfortunately, the procedure suggested by Nayak requires an external boiler of adequate volume and power and the related connecting piping to the reactor vessel, while the procedure suggested by Jiang and coauthors requires an additional system for the nitrogen storage and the related connecting piping to the reactor vessel. The external pressurization does not accomplish to the requirements of simplicity that are at the very base of natural circulation BWRs design and it is thus not recommendable. Cheung and Rao [3] suggested a start-up procedure in which the reactor is first filled with water at 80 deg. C at a pressure of 0.55 bar. The reactor is made critical and is pressurized in conditions of single-phase circulation up to a pressure of 63 bar. At this pressure a sudden transition to two-phase operation is achieved by opening the MSIVs (Main Steam Isolation

  3. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph, E-mail: joseph.nielsen@inl.gov [Idaho National Laboratory, 1955 N. Fremont Avenue, P.O. Box 1625, Idaho Falls, ID 83402 (United States); University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Tokuhiro, Akira [University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Hiromoto, Robert [University of Idaho, Department of Computer Science, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Tu, Lei [University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States)

    2015-12-15

    state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This paper presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. In order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

  4. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    International Nuclear Information System (INIS)

    Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; Tu, Lei

    2015-01-01

    state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This paper presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. In order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

  5. Expected returns from a tax on nuclear fuel elements in the context of longer service lives of German nuclear power plants

    International Nuclear Information System (INIS)

    Kondziella, Hendrik; Bruckner, Thomas

    2010-01-01

    To what extent will the fuel element tax introduced by the German government in combination with the longer service life of nuclear power stations reduce the profits of public utilities? A qualitative assessment suggests that the tax will not equal the full profits. Using an electricity market model, various scenarios can be calculated for an eight-year prolongation of the residual service life of existing nuclear power plants. (orig.)

  6. Plant specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events

    International Nuclear Information System (INIS)

    2011-01-01

    The German Parliament requested (17 March 2011) a comprehensive inspection of German nuclear power plants. For this purpose independent expert commissions should perform a new risk analysis of all German NPPS and nuclear installations with respect to the lessons learned from the Fukushima (Japan) events and other extraordinary damage scenarios. The Reactor safety commission (RSK) was assigned by the German Bundesamt fuer Strahlenschutz to develop a catalogue of requirements for this safety inspection. The contribution summarizes the required inspection volume (status 30.03.2011) including the following events: natural events like earth quakes, floods, weather-based consequences and possible superposition. Additionally the following assumptions have to be considered: event independent postulated common failures or systematic faults, station blackout larger than 2 hours, long-term failure of the auxiliary cooling water supply; aggravating boundary conditions for the performance of emergency measures (non-availability of power supply), hydrogen generation and detonation hazard, restricted personnel availability, non-accessibility due to high radiation levels, impeded technical support from outside. (orig.)

  7. Analysis of human factor aspects in connection with available incident reports obligatorily reported by German nuclear power plants

    International Nuclear Information System (INIS)

    Wilpert, B.; Freitag, M.; Miller, R.

    1993-01-01

    Goal of the present study is the analysis of human factor aspects in connection with available incident reports obligatorily reported by German nuclear power plants. Based on psychological theories and empirical studies this study develops a classification scheme which permits the identification of foci of erroneous human actions. This classification scheme is applied to a selection of human factor relevant incidents by calculating frequencies of the occurrence of human error categories. The results allow insights into human factor related problem areas. (orig.) [de

  8. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Peng, C.M.; Maly, J.

    1988-01-01

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  9. Subchannel analysis program for boiling water reactor fuel bundles based on five conservation equations of two-phase flow

    International Nuclear Information System (INIS)

    Bessho, Y.; Uchikawa, S.

    1985-01-01

    A subchannel analysis program, MENUETT, is developed for evaluation of thermal-hydraulic characteristics in boiling water reactor fuel bundles. This program is based on five conservation equations of two-phase flow with the drift-flux correlation. The cross flows are calculated separately for liquid and vapor phases from the lateral momentum conservation equation. The effects of turbulent mixing and void drift are accounted for in the program. The conservation equations are implicitly differentiated with the convective terms by the donor-cell method, and are solved iteratively in the axial and lateral directions. Data of the 3 X 3 rod bundle experiments are used for program verification. The lateral distributions of equilibrium quality and mass flow rate at the bundle exit calculated by the program compare satisfactorily with the experimental results

  10. A study of thermal-hydraulic requirements for increasing the power rates for natural-circulation boiling water reactors

    International Nuclear Information System (INIS)

    Yasuo, A.; Inada, F.; Hidaka, M.

    1992-01-01

    In this paper, the feasibility of higher power rates for natural-circulation boiling water reactors (BWRs) is studied with the objective of examining the flexibility of the plant power rate in constructing such plants to cope with the increasing demand for electricity. By applying existing one-dimensional design codes, the riser heights necessary to meet two major thermal-hydraulic requirements, i.e., critical power and core stability, are systematically calculated. Several restrictions on the maximum diameter and height of the pressure vessel are also considered because these restrictions could make construction impossible or drastically increase the construction costs. A very simple map of the dominant parameters for higher power rates is obtained. It is concluded that natural-circulation BWRs of >1000 MW (electric) will be feasible within the restrictions considered here

  11. Analytic solution to verify code predictions of two-phase flow in a boiling water reactor core channel

    International Nuclear Information System (INIS)

    Chen, K.F.; Olson, C.A.

    1983-01-01

    One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction

  12. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  13. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-01-01

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to: (1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, (2) assess the RELAP5 and TRACE computer code against the experimental data, and (3) develop mathematical model and heat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal-hydraulic codes assessment

  14. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  15. Method for optimum determination of adjustable parameters in the boiling water reactor core simulator using operating data on flux distribution

    International Nuclear Information System (INIS)

    Kiguchi, T.; Kawai, T.

    1975-01-01

    A method has been developed to optimally and automatically determine the adjustable parameters of the boiling water reactor three-dimensional core simulator FLARE. The steepest gradient method is adopted for the optimization. The parameters are adjusted to best fit the operating data on power distribution measured by traversing in-core probes (TIP). The average error in the calculated TIP readings normalized by the core average is 0.053 at the rated power. The k-infinity correction term has also been derived theoretically to reduce the relatively large error in the calculated TIP readings near the tips of control rods, which is induced by the coarseness of mesh points. By introducing this correction, the average error decreases to 0.047. The void-quality relation is recognized as a function of coolant flow rate. The relation is estimated to fit the measured distributions of TIP reading at the partial power states

  16. An overview of the U.S. Department of Energy Experimental Boiling Water Reactor Decontamination and Decommissioning Project

    International Nuclear Information System (INIS)

    Murphie, W.E.; Mckernan, M.L.

    1991-01-01

    This paper provides an overview of the U.S. Department of Energy's (DOE) Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D and D) Project. Physical decommissioning work started in 1986 and is scheduled for completion in 1994. The project total estimated cost is 14.3 million (1990, U.S.) dollars. The reactor pressure vessel will be removed by segmentation. Another notable project feature is that D and D operations were planned for and carried out with a small work force comprised of four to six D and D laborers, one or two health physics technicians, an engineer, and a project manager. When the D and D work is completed the facility will be recycled for other productive uses. (author)

  17. Cycle studies: material balance estimation in the domain of pressurized water and boiling water reactors. Experimental qualification

    International Nuclear Information System (INIS)

    Chabert, Christine

    1994-01-01

    This study is concerned with the physics of the fuel cycle the aim being to develop and make recommendations concerning schemes for calculating the neutronics of light water reactor fuel cycles. A preliminary study carried out using the old fuel cycle calculation scheme APOLLO1- KAFKA and the library SERMA79 has shown that for the compositions of totally dissolved assemblies from Pressurized Water Reactors (type 17*17) and also for the first time, for Boiling Water Reactor assemblies (type 8*8), the differences between calculation and measurement are large and must be reduced. The integration of the APOLLO2 neutronics code into the fuel cycle calculation scheme improves the results because it can model the situation more precisely. A comparison between APOLLO1 and APOLLO2 using the same options, demonstrated the consistency of the two methods for PWR and BWR geometries. Following this comparison, we developed an optimised scheme for PWR applications using the library CEA86 and the code APOLLO2. Depending on whether the information required is the detailed distribution of the composition of the irradiated fuel or the average composition (estimation of the total material balance of the fuel assembly), the physics options recommended are different. We show that the use of APOLLO2 and the library CEA86 improves the results and especially the estimation of the Pu 239 content. Concerning the Boiling Water Reactor, we have highlighted the need to treat several axial sections of the fuel assembly (variation of the void-fraction, heterogeneity of composition). A scheme using Sn transport theory, permits one to obtain a better coherence between the consumption of U 235 , the production of plutonium and burnup, and a better estimation of the material balance. (author) [fr

  18. In-core power sharing and fuel requirement study for a decommissioning Boiling Water Reactor using the linear reactivity model

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shung-Jung; Kuo, Weng-Sheng

    2014-01-01

    Highlights: • Linear reactivity model (LRM) was modified and applied to Boiling Water Reactor. • The power sharing and fuel requirement study of the last cycle and two cycles before decommissioning was implemented. • The loading pattern design concept for the cycles before decommissioning is carried out. - Abstract: A study of in-core power sharing and fuel requirement for a decommissioning BWR (Boiling Water Reactor) was carried out using the linear reactivity model (LRM). The power sharing of each fuel batch was taken as an independent variable, and the related parameters were set and modified to simulate actual cases. Optimizations of the last cycle and two cycles before decommissioning were both implemented; in the last-one-cycle optimization, a single cycle optimization was carried out with different upper limits of fuel batch power, whereas, in the two-cycle optimization, two cycles were optimized with different cycle lengths, along with two different optimization approaches which are the simultaneous optimization of two cycles (MO) and two successive single-cycle optimizations (SO). The results of the last-one-cycle optimization show that it is better to increase the fresh fuel power and decrease the thrice-burnt fuel power as much as possible. It also shows that relaxing the power limit is good to the fresh fuel requirement which will be reduced under lower power limit. On the other hand, the results of the last-two-cycle (cycle N-1 and N) optimization show that the MO is better than SO, and the power of fresh fuel batch should be decreased in cycle N-1 to save its energy for the next cycle. The results of the single-cycle optimization are found to be the same as that in cycle N of the multi-cycle optimization. Besides that, under the same total energy requirement of two cycles, a long-short distribution of cycle length design can save more fresh fuel

  19. The 2002 amendment to the German atomic energy act concerning the phase-out of nuclear power

    International Nuclear Information System (INIS)

    Vorwerk, A.

    2002-01-01

    The phase-out of the use of nuclear power for electricity production has now been legally regulated by the 2002 Atomic Energy Act, based on the Agreement between the German Government and the energy utilities. The provisions of this Act comply with constitutional and European law, and take account of Germany's international commitments. The new 2002 Atomic Energy Act is supplemented by additional steps towards the phase-out, in particular in the area of nuclear disposal. These steps are being taken primarily within the framework of a planned national disposal plan and a procedure to be enveloped for the selection of a location for a final disposal site for radioactive wastes. The key task for the Laender authorities and the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety continues to be to ensure that operators of nuclear power plants comply with a high standard of safety during the residual operating periods of their plants. (author)

  20. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riley, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schroeder, John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Aldrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maljovec, Dan [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Bie [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pascucci, Valerio [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  1. Advanced Best-Estimate Methodologies for Thermal-Hydraulics Stability Analyses with TRACG code and Improvements on Operating Boiling Water Reactors

    International Nuclear Information System (INIS)

    Vedovi, J.; Trueba, M.; Ibarra, L; Espino, M.; Hoang, H.

    2016-01-01

    In recent years GE Hitachi has introduced two advanced methodologies to address the thermal-hydraulics instabilities in Boiling Water Reactors (BWRs); the “Detect and Suppress Solution - Confirmation Density (DSS-CD)” and the “GEH Simplified Stability Solution (GS3).” These two methodologies are based on Best-Estimate Plus Uncertainty (BEPU) analyses and provide significant improvement on safety, plant maneuvering and fuel economics with respect to existing solutions. DSS-CD and GS3 solutions have been recently approved by the United States Nuclear Regulatory Commission. This paper describes the main characteristics of these two stability methodologies and shares the experience of their recent implementation in operating BWRs. The BEPU approach provided a much deeper understanding of the parameters affecting instabilities in operating BWRs and allowed for better calculation of plant setpoints by improving plant manoeuvring restrictions and reducing manual operator actions. DSS-CD and GS3 methodologies are both based on safety analyses performed with the best-estimate system code TRACG. The assessment of uncertainty is performed following the Code Scaling, Applicability and Uncertainty (CSAU) methodology documented in NUREG/CR-5249. The two solutions have been already implemented in a combined 18 BWR units with 7 more units in the process of transitioning. The main results demonstrate a significant decrease (>0.1) in the stability based Operating Limit Minimum Critical Power Ratio (OLMCPR), which possibly results in significant fuel savings and the increase in allowable stability plant setpoints that address instability events such as the one occurred at the Fermi 2 plant in 2015 and can help prevent unnecessary Scrams. The paper also describes the advantages of reduced plant manoeuvring as a result to transitioning to these solutions; in particular the history of a BWR/6 transition to DSS-CD is discussed.

  2. The characteristic features of the German and the U.S. American law governing the licensing of nuclear installations

    International Nuclear Information System (INIS)

    Heitsch, C.

    1993-01-01

    The author's comparative analysis of the German and the U.S. American legislation is intended to exhibit the specific structures of the two regulatory systems, and their performance regarding the intended purpose of providing the best possible protection against the risks posed by nuclear energy. It is further analysed whether instruments or structures of the U.S. American law can be adopted by the German law and applied to solving problems arising in Germany. In the selfcontained review of the German Atomic Energy Act and the nuclear licensing provisions, emphasis is placed on the development of judicial control in Germany, as the court decisions have a major impact on accepted practice in Germany. This is followed by a self-contained review of the licensing procedure in the U.S.A., with a focus on the right to be heard of the general public, and on the system of rulemaking powers by the state commissions and agencies. The author finally explains the most significant differences of the two regulatory systems and draws some conclusions as to the underlying legal policy. A concluding analysis examines the question of whether the recent judgments by the Federal Administrative Court, relating to judicial control and the binding effect of substantiating administrative regulations, are justifiable from the dogmative point of views. (orig.) [de

  3. Harmonization of French and German calculation procedures for atmospheric dispersion following accidental releases from nuclear power plants

    International Nuclear Information System (INIS)

    Crabol, B.; Romeo, E.; Nester, K.

    1992-01-01

    In case of an accident in a nuclear power plant near the French-German border different schemes for dispersion calculations in both countries will currently be applied. An intercomparison of these schemes initiated from the German-French Commission for the safety of nuclear installations (DFK) revealed in some meteorological situations large differences in the resulting concentrations for radionuclides. An ad hoc working group was installed by the DFK with the mandate to analyse the reasons for the different model results and also to consider new theoretical concepts. The working group has agreed to apply a Gaussian puff model for emergency response calculations. The results of the model based on turbulence parameterization via similarity approach or spectral theory - have been compared with tracer experiments for different emission heights and atmospheric stability regimes. As a reference the old modelling approaches have been included in the study. The simulations with the similarity approach and the spectral theory show a slightly better agreement to the measured concentration data than the schemes used in the past. Instead of diffusion categories both new approaches allow a continuous characterization of the atmospheric dispersion conditions. Because the spectral approach incorporates the sampling time of the meteorological data as an adjustable parameter thereby offering the possibility to adjust the dispersion model to different emission scenarios this turbulence parameterization scheme will be foreseen as the basis for a joint French-German puff model

  4. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    International Nuclear Information System (INIS)

    1993-01-01

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval

  5. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  6. Checking technical measurements on climatic data during sand blasting and spraying work in the condensation chamber of the boiling water reactor Gundremmingen

    International Nuclear Information System (INIS)

    Rausch, D.; Unte, U.

    1986-01-01

    During sand blasting and spraying work in the condensation chambers of boiling water reactors prescribed climatic data must be adhered to. For this purpose temporary air conditioners are used. The technical measurement examination here should provide information as to whether the air conditioners used were to fulfill the parameter curve specifications. (orig.) [de

  7. Verification of HELIOS-MASTER system through benchmark of Halden boiling water reactor (HBWR)

    International Nuclear Information System (INIS)

    Kim, Ha Yong; Song, Jae Seung; Cho, Jin Young; Kim, Kang Seok; Lee, Chung Chan; Zee, Sung Quun

    2004-01-01

    To verify the HELIOS-MASTER computer code system for a nuclear design, we have been performed benchmark calculations for various reactor cores. The Halden reactor is a boiling, heavy water moderated reactor. At a full power of 18-20MWt, the moderator temperature is 240 .deg. C and the pressure is 33 bar. This study describes the verification of the HELIOS-MASTER computer code system for a nuclear design and the analysis of a hexagonal and D 2 O moderated core through a benchmark of the Halden reactor core. HELIOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves the neutronics model with the TPEN (Triangle based Polynomial Expansion Nodal) method for a hexagonal geometry

  8. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

    ... Environmental Impact: Availability IX. Paperwork Reduction Act Statement X. Regulatory Analysis XI. Regulatory... designs to perform a design-specific assessment of the effects of the impact of a large commercial.... ABWR; one commenter, GE Hitachi Nuclear Energy (GEH), was against the proposed amendment to the U.S...

  9. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... exemptions from specific emergency planning requirements of part 50 of Title 10 of the Code of Federal... exempting the facility from the emergency planning requirements will not have any adverse environmental..., Office of Federal and State Materials and Environmental Management Programs, U.S. Nuclear Regulatory...

  10. Anticipatory opt-out of nuclear power at German universities and research centers? Employment and education opportunities in the nuclear field

    International Nuclear Information System (INIS)

    Fritz, P.; Closs, K.D.; Kuczera, B.; Erlenwein, P.; Langetepe, G.; Weiss, F.P.; Wolfert, K.

    2001-01-01

    The energy policy objectives of the present German federal government include the regulated, not-to-be-indemnified termination of the use of nuclear power for electricity generation in Germany. This objective was initialed in an agreement between the federal government and the operators of nuclear power plants in June 2000. The criterion it contains for the residual life of existing nuclear power plants are the 'residual electricity volumes' to be generated by existing plants. According to present data, the first nuclear power plant would be decommissioned in late 2002, the last one, in early 2021. A report by an evaluating committee established by the competent German Federal Ministry of Economics and Technology published in January 2000 indicates that, even in the light of this medium-term perspective for the use of nuclear power for electricity generation, sufficient competence in terms of manpower in this field must be maintained and promoted in all areas of research, industry, expert consultant organizations, and regulatory authorities far beyond 2021. This makes government support indispensable in the light of the political criteria requiring the preservation of competence in this sector. The 'Kompetenzverbund Kerntechnik' (Association for Nuclear Technology Competence) was established within the framework of the Energy Research Division of the Hermann von Helmholtz-Gemeinschaft Deutscher Forschungszentren. Its mandate outlines its objectives as follows: 'Irrespective of political decisions establishing criteria for terminating the use of nuclear power in Germany, it is necessary that competence in nuclear safety be preserved for the next few decades. Only this will ensure that the government's provident duties can be fulfilled, and the safety of nuclear installations and waste and spent fuel management pathways can be ensured in accordance with the international state of the art'. (orig.) [de

  11. Development of natural circulation small and medium sized boiling water reactor: HSBWR-600

    International Nuclear Information System (INIS)

    Miki, Minoru; Horiuchi, Tetsuo; Yoshimoto, Yuichiro; Sumida, Isao; Murase, Michio; Akita, Minoru; Niino, Tsuyoshi

    1988-01-01

    In nuclear power generation, the development of large reactors has been promoted as the main energy source in Japan. However, world economy entered low growth age, and the growth of electric power demand slowed down. Accordingly, attention has been paid to the medium and small reactors that can cope with whatever needs by serializing their types in addition to the nuclear power plants of medium output matching to electric power demand. In order to cope with these new needs, the economical efficiency of medium and small reactors must be as close as possible to that of large reactors, and as the countermeasures to the demerits due to small size, those must be made into the plants having simplified systems and the safety easily acceptable to public. Hitachi Ltd. plans to develop the natural circulation type medium and small BWRs of 600 NWe output class, HSBWR-600, on the basis of the nuclear power plant technology based on the rich results of design and operation of BWRs obtained so far, and to rank them as one of the BWR series. The target of their development design, the circumstance of their development, the core design and the thermo-hydraulic characteristics, the reactor pressure vessel and in-core structures, the safety design, system design, building layout and the evaluation are reported. (Kako, I.)

  12. A model of Alto Lazio boiling water reactor using the LEGO code balance of plant simulation

    International Nuclear Information System (INIS)

    Spelta, S.; Garbossa, G.B.

    1989-01-01

    An extensive effort has been made at the Italian National Electricity Board (ENEL) to construct and validate a LEGO model capable of simulating the operational transients of the Alto Lazio Nuclear Station, a two twin-units site with BWR/6 class reactors, rated at 2894 MWt and with Mark III containment. LEGO is a modular package developed at the Research and development Department of the Italian National Electricity Board (CRA-ENEL) for computer aided modeling of fossil-fired and nuclear steam power plants. In this paper a system analysis model capable of describing steady-state and transient performance of the Balance of Plant (BOP) of the Alto Lazio Power Station is presented. This is one of two companion papers devoted to the description of the overall plant model including both the Nuclear Steam Supply System (NSSS) and the BOP. In the paper, after a brief summary of the main LEGO characteristics, a description of the BOP lay-out is presented. The overall model, which has been set-up, including control systems and automation, is very detailed and consists of almost 2000 differential or algebraic equations. After a brief description of the mathematical model, two significant transients obtained using the overall model are presented and discussed

  13. Application of the accident consequences model of the German risk study to assessments of accident risks in different types of nuclear power plants

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Bayer, A.

    1982-01-01

    Within the scope of the 'German Risk Study for Nuclear Power Plants' (Phase A) the accident consequence model UFOMOD was developed in the Karlsruhe Nuclear Research Center. This model originally developed for pressurized water reactors has now been extended in order to obtain results about accidental releases of activity from fast breeder and high-temperature reactors, too. (RW) [de

  14. Technical evaluation of RETS-required reports for the La Crosse boiling water reactor

    International Nuclear Information System (INIS)

    Magleby, E.H.; Young, T.E.

    1985-01-01

    A review of the reports required by federal regulations and the plant-specific Radiological Effluent Technical Specifications (RETS) for operations conducted during 1983 was performed. The periodic reports reviewed were the Annual Radiological Environmental Operating Report for 1983 and the Semiannual Radioactive Effluent Release Reports for 1983. The principal review guidelines were the plant-specific RETS, NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants'', and NRC Guidance on the Review of the Process Control Programs. The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines. 5 refs

  15. RETRAN operational transient analysis of the Big Rock Point plant boiling water reactor

    International Nuclear Information System (INIS)

    Sawtelle, G.R.; Atchison, J.D.; Farman, R.F.; VandeWalle, D.J.; Bazydlo, H.G.

    1983-01-01

    Energy Incorporated used the RETRAN computer code to model and calculate nine Consumers Power Company Big Rock Point Nuclear Power Plant transients. RETRAN, a best-estimate, one-dimensional, homogeneous-flow thermal-equilibrium code, is applicable to FSAR Chapter 15 transients for Conditions 1 through IV. The BWR analyses were performed in accordance with USNRC Standard Review Plan criteria and in response to the USNRC Systematic Evaluation Program. The RETRAN Big Rock Point model was verified by comparison to plant startup test data. This paper discusses the unique modeling techniques used in RETRAN to model this steam-drum-type BWR. Transient analyses results are also presented

  16. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-09-30

    The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask, in part by increasing the efficiency of internal conduction pathways, and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above- and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an above-ground configuration.

  17. Minimizing the power peaking factor of fuel lattices using factors of group for boiling water reactors

    International Nuclear Information System (INIS)

    Guzman, J. R.; Longoria, L. C.; De la Cruz, E.; Arredondo, C.

    2010-10-01

    A method to design the distribution and composition of nuclear fuel for the array of fuel rods in a lattice for BWRs is presented in this work. The aim of the method is to minimize the power peaking factor until an adequate value is reached. Also, this method uses a few calculations of lattice. The method is based on the classification of the fuel rods in two groups: the group of fuel rods with the higher power level (group pow ), and the other group of fuel rods (no-group pow ). The enrichment of 235 U of each fuel rod of the group pow is multiplied by a factor called group fissile factor (f group ), and the enrichment of 235 U of each fuel rod of the no-group pow is multiplied by a factor called no-group fissile factor (f no-group ). These factors are fitted so that the power peaking factor is minimized. The importance of the method with the use of these two factors is applied to the design of a fuel lattice for BWRs as the Laguna Verde nuclear power plant. The calculations of lattice are made by means of the Helios code. (Author)

  18. Improvement of daily load-following operation for boiling water reactors

    International Nuclear Information System (INIS)

    Kiguchi, Takashi; Kurihara, Kunitoshi; Sakurai, Mikio; Joge, Toshio; Asami, Kazuo.

    1980-01-01

    Recently, with the increase of the proportion of nuclear power generation to the total amount of power generation of electric power systems, the needs of daily load-following operation of nuclear power stations have heightened, accordingly the study on the method of daily load-following operation has been carried out for BWRs. In this study, by the combined use of the flow rate control of core coolnat being operated easily and the operation of control rods, the BWR system with the daily load-following performance of 100% power output in daytime and 50% power output at night was the target of development. For the purpose, the change of core characteristics during load-following was grasped analytically, and the range of load change was investigated. At the same time, as the first stage of developing operation control and monitoring system, the reactor output-adjusting device which makes generator output automatically follow the target load change pattern by the flow rate control of core coolnat, and the equipment for monitoring core performance on line were developed. The analysis of the method of daily load-following operation in present-day BWRs, the study on the improvement of load-following operation performance, the reactor output-adjusting device are described. (Kako, I.)

  19. The current situation of waste management at German nuclear power plants

    International Nuclear Information System (INIS)

    Hottenrott, G.; Mirschinka, V.

    2002-01-01

    An overview is given about the main points of the agreement between the four utility companies and the German government on June 11, 2001, concerning waste management and disposal. The current waste management situation is compared with the effects of the agreement on the future development of the German waste management concept. Major innovations are the discontinuation of reprocessing, the establishment of local intermediate storage facilities for fuel elements and the suspension of the exploration works conducted on the salt dome at Gorleben. (orig.)

  20. Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

    International Nuclear Information System (INIS)

    Hopkins, G.C.

    1982-05-01

    This report documents benchmark evaluations to test thorium lattice predictive methods and neutron cross sections against available data and summarizes specific evaluations of the once-through thorium cycle when applied to the Browns Ferry-3 BWR. It was concluded that appreciable uncertainties in thorium cycle nuclear data cloud the ability to reliably predict the fuel cycle performance and that power reactor irradiations of ThO 2 rods in BWRs are desirable to resolve uncertainties. Benchmark evaluations indicated that the ENDF/B-IV data used in the evaluations should cause an underprediction of U-233/ThO 2 fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative