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Sample records for bwr power plants

  1. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

  2. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

    Andrade, G.G. de.

    1973-01-01

    First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

  3. Development of a computerized operator support system for BWR power plant

    International Nuclear Information System (INIS)

    Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

    1985-01-01

    A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

  4. Development and recent trend of design of BWR nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kani, J [Tokyo Shibaura Electric Co. Ltd., Kawasaki, Kanagawa (Japan)

    1977-11-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation.

  5. Development and recent trend of disign of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Kani, Jiro

    1977-01-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation. (Wakatsuki, Y.)

  6. Operation status display and monitoring system for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Wakabayashi, Yasuo; Hayakawa, Hiroyasu; Kawamura, Atsuo; Kaneda, Mitsunori.

    1982-01-01

    Lately, the development of the system has been made for BWR plants, which monitors the operating status not only in normal operation but also in abnormal state and also for plant safety. Recently, the improvement of man-machine interface has been tried through the practical use of technique which displays data collectively on a CRT screen relating them mutually. As one of those results, the practical use of an electronic computer and color CRT display for No. 1 unit in the Fukushima No. 2 Nuclear Power Station (2F-1), Tokyo Electric Power Co., is described. Also, new centralized control panels containing such systems were used for the 1100 MWe BWR nuclear power plants now under construction, No. 3 unit of the Fukushima No. 2 Power Station and No. 1 unit of Kashiwazaki-Kariwa Nuclear Power Station (2F-3 and K-1, respectively). The display and monitoring system in 2F-1 plant is the first one in which a computer and color CRTs were practically employed for a BWR plant in Japan, and already in commercial operation. The advanced operating status monitoring system, to which the result of evaluation of the above system was added, was incorporated in the new centralized control panels presently under production for 2F-3 and K-1 plants. The outline of the system, the functions of an electronic computer, plant operating status monitor, surveillance test guide, the automation of plant operation and auxiliary operation guide are reported for these advanced monitoring system. It was confirmed that these systems are useful means to improve the man-machine communication for plant operation minitoring. (Wakatsuki, Y.)

  7. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  8. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake

    1995-01-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

  9. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

    1995-03-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

  10. Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

    1986-01-01

    The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

  11. Development of next BWR plant

    International Nuclear Information System (INIS)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

    1995-01-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

  12. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1995-04-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

  13. Description of the power plant model BWR-plasim outlined for the Barsebaeck 2 plant

    International Nuclear Information System (INIS)

    Christensen, P. la Cour.

    1979-08-01

    A description is given of a BWR power plant model outlined for the Barsebaeck 2 plant with data placed at our disposal by the Swedish Power Company Sydkraft A/B. The basic operations are derived and simplifications discussed. The model is implemented with a simulation system DYSYS which assures reliable solutions and easy programming. Emphasis has been placed on the models versatility and flexibility so new features are easy to incorporate. The model may be used for transient calculations for both normal plant conditions and for abnormal occurences as well as for control system studies. (author)

  14. Method of operating BWR type power plants

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To improve the operation efficiency of BWR type reactors by reducing the time from the start-up of the reactor to the start-up of the turbine and electrical generator, as well as decrease the pressure difference in each of the sections of the pressure vessel to thereby extend its life span. Method: The operation comprises switching the nuclear reactor from the shutdown mode to the start-up mode, increasing the reactor power to a predetermined level lower than a rated power while maintaining the reactor pressure to a predetermined level lower than a rated pressure, starting up a turbine and an electrical generator in the state of the predetermined reactor pressure and the reactor power to connect the electrical generator to the power transmission system and, thereafter, increasing the reactor pressure and the reactor power to the predetermined rated pressure and rated power respectively. This can shorten the time from the start-up of the reactor to the start of the power transmission system, whereby the operation efficiency of the power plant can be improved. (Moriyama, K.)

  15. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  16. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  17. 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)

  18. Trend of field data on pipe wall thinning for BWR power plants

    International Nuclear Information System (INIS)

    Hakii, Junichi; Hiranuma, Naoki; Hidaka, Akitaka

    2009-01-01

    Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes and Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plans from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulence and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement). (author)

  19. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  20. LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

    2005-01-01

    characteristics. In this paper, related publications will be reviewed to enhance the understanding of parametric effects to the system stability. Subsequently, core stability analysis for the Kuosheng Nuclear Power Plant (BWR6) will be conducted, and then parallel verification against vendor's results will be performed also. We will summarize our efforts in using LAPUR5 methodologies for BWR6 core stability analysis and share our experiences among international communities. (authors)

  1. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  2. Laguna Verde nuclear power plant: an experience to consider in advanced BWR design

    International Nuclear Information System (INIS)

    Fuentes Marquez, L.

    2001-01-01

    Laguna Verde is a BWR 5 containment Mark II. Designed by GE, two external re-circulation loops, each of them having two speed re-circulation pump and a flow control valve to define the drive flow and consequently the total core flow an power control by total core flow. Laguna Verde Design and operational experience has shown some insights to be considering in design for advanced BRW reactors in order to improve the potential of nuclear power plants. NSSS and Balance of plant design, codes used to perform nuclear core design, margins derived from engineering judgment, at the time Laguna Verde designed and constructed had conducted to have a plant with an operational license, generating with a very good performance and availability. Nevertheless, some design characteristics and operational experience have shown that potential improvements or areas of opportunity shall be focused in the advanced BWR design. Computer codes used to design the nuclear core have been evolved relatively fast. The computers are faster and powerful than those used during the design process, also instrumentation and control are becoming part of this amazing technical evolution in the industry. The Laguna Verde experience is the subject to share in this paper. (author)

  3. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Monta, K.; Itoh, J.

    1992-01-01

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  4. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  5. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  6. Distributed control and data processing system with a centralized database for a BWR power plant

    International Nuclear Information System (INIS)

    Fujii, K.; Neda, T.; Kawamura, A.; Monta, K.; Satoh, K.

    1980-01-01

    Recent digital techniques based on changes in electronics and computer technologies have realized a very wide scale of computer application to BWR Power Plant control and instrumentation. Multifarious computers, from micro to mega, are introduced separately. And to get better control and instrumentation system performance, hierarchical computer complex system architecture has been developed. This paper addresses the hierarchical computer complex system architecture which enables more efficient introduction of computer systems to a Nuclear Power Plant. Distributed control and processing systems, which are the components of the hierarchical computer complex, are described in some detail, and the database for the hierarchical computer complex is also discussed. The hierarchical computer complex system has been developed and is now in the detailed design stage for actual power plant application. (auth)

  7. Operational experience of human-friendly control and instrumentation systems for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Makino, M.; Watanabe, T.; Suto, O.; Asahi, R.

    1987-01-01

    In recent BWR nuclear power plants in Japan, an advanced centralized monitoring and control system PODIA (Plant Operation by Displayed Information and Automation), which incorporates many operator aid functions, has been in operation since 1985. Main functions of the PODIA system as a computerized operator aid system are as follows. CRT displays for plant monitoring. Automatic controls and operation guides for plant operation. Stand-by status monitoring for engineered safety features during normal operation. Surveillance test procedure guides for engineered safety features. Integrated alarm display. The effectiveness of these functions have been proved through test and commercial operation. It has been obtained that operators have preferred PODIA much more than conventional monitoring and control systems

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

    International Nuclear Information System (INIS)

    1992-09-01

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

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

    International Nuclear Information System (INIS)

    1992-09-01

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

  10. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

    2008-01-01

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  11. Development and Evaluation of cooperative control system for an HVDC transmission system connected with an isolated BWR power plant

    International Nuclear Information System (INIS)

    Horiuchi, Susumu; Hara, Tsukusi; Matori, Iwao; Hirayama, Kaiichirou.

    1987-01-01

    In the cooperative control system developed for an HVDC transmission system connected with an isolated BWR power plant, the equilibrium state between power plant output and DC transmission power is examined by way of the detection of the generator frequency. And, thereby start-up and shutdown of the DC system and controlling of the transmission power are made, so that the signal transmission with the power plant becomes unnecessary, enabling the easy cooperative operation. In order to investigate validity of this control system, various digital simulation and simulator test with the control system were carried out. In this way, behavior of the power plant and stability of the DC transmission system were evaluated in the connection to the DC system at power plant start-up, follow of the transmission power in change of the power plant output and in various system failures. (Mori, K.)

  12. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

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

    International Nuclear Information System (INIS)

    1992-09-01

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

  14. Parameter identification of a BWR nuclear power plant model for use in optimal control

    International Nuclear Information System (INIS)

    Volf, K.

    1976-02-01

    The problem being considered is the modeling of a nuclear power plant for the development of an optimal control system of the plant. Current system identification concepts, combining input/output information with a-priori structural information are employed. Two of the known parameter identification methods i.e., a least squares method and a maximum likelihood technique, are studied as ways of parameter identification from measurement data. A low order state variable stochastic model of a BWR nuclear power plant is presented as an application of this approach. The model consists of a deterministic and a noise part. The deterministic part is formed by simplified modeling of the major plant dynamic phenomena. The moise part models the effects of input random disturbances to the deterministic part and additive measurement noise. Most of the model parameters are assumed to be initially unknown. They are identified using measurement data records. A detailed high order digital computer simulation is used to simulate plant dynamic behaviour since it is not conceivable for experimentation of this kind to be performed on the real nuclear power plant. The identification task consists in adapting the performance of the simple model to the data acquired from this plant simulation ensuring the applicability of the techniques to measurement data acquired directly from the plant. (orig.) [de

  15. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  16. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  17. Experience and development of on-line BWR surveillance system at Onagawa nuclear power station unit-1

    International Nuclear Information System (INIS)

    Kishi, A.; Chiba, K.; Kato, K.; Ebata, S.; Ando, Y.; Sakamoto, H.

    1986-01-01

    ONAGAWA nuclear power station Unit-1 (Tohoku Electric Power Co.) is a BWR-4 nuclear power station of 524 MW electric power which started commercial operation in June 1984. To attain high reliability and applicability for ONAGAWA-1, Tohoku Electric Power Co. and Toshiba started a Research and Development project on plant surveillance and diagnosis from April 1982. Main purposes of this project are to: (1) Develop an on-line surveillance system and acquire its operating experience at a commercial BWR, (2) Assist in plant operation and maintenance by data acquisition and analysis, (3) Develop a new technique for plant surveillance and diagnosis. An outline of the project, operating experience gained from the on-line surveillance system and an introduction to new diagnosis techniques are reported in this paper. (author)

  18. Quantitative evaluation for training results of nuclear plant operator on BWR simulator

    International Nuclear Information System (INIS)

    Sato, Takao; Sato, Tatsuaki; Onishi, Hiroshi; Miyakita, Kohji; Mizuno, Toshiyuki

    1985-01-01

    Recently, the reliability of neclear power plants has largely risen, and the abnormal phenomena in the actual plants are rarely encountered. Therefore, the training using simulators becomes more and more important. In BWR Operator Training Center Corp., the training of the operators of BWR power plants has been continued for about ten years using a simulator having the nearly same function as the actual plants. The recent high capacity ratio of nuclear power plants has been mostly supported by excellent operators trained in this way. Taking the opportunity of the start of operation of No.2 simulator, effort has been exerted to quantitatively grasp the effect of training and to heighten the quality of training. The outline of seven training courses is shown. The technical ability required for operators, the items of quantifying the effect of training, that is, operational errors and the time required for operation, the method of quantifying, the method of collecting the data and the results of the application to the actual training are described. It was found that this method is suitable to quantify the effect of training. (Kako, I.)

  19. Recent operating experience during startup testing at latest 1100 MWe BWR-5 nuclear power plants

    International Nuclear Information System (INIS)

    Tanabe, Akira; Tateishi, Mizuo; Kajikawa, Makoto; Hayase, Yuichi.

    1986-01-01

    In June and September 1985, the latest two 1100 Mwe BWR-5 nuclear power plants started commercial operation about ten days earlier than initially expected without any unscheduled shutdown. These latest plants, 2F-3 and K-1, are characterized by an improved core with new 8 x 8 fuel assemblies, highly reliable control systems, advanced control room system and turbine steam full bypass system for full load rejection (2F3). This paper describes the following operating experiences gained during their startup testing. 1) Continuous operation at full load rejection. 2) Stable operation at natural circulating flow condition. 3) 31 and 23 hour short time start up operation. 4) 100-75-100 %, 1-8-1-14 hours daily load following operation. (author)

  20. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  1. ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics

    International Nuclear Information System (INIS)

    Fuller, L.C.; Myers, M.L.

    1975-01-01

    1 - Description of problem or function: ORCOST2 estimates the cost of electrical energy production from single-unit steam-electric power plants. Capital costs and operating and maintenance costs are calculated using base cost models which are included in the program for each of the following types of plants: PWR, BWR, HTGR, coal, oil, and gas. The user may select one of several input/output options for calculation of capital cost, operating and maintenance cost, levelized energy costs, fixed charge rate, annual cash flows, cumulative cash flows, and cumulative discounted cash flows. Options include the input of capital cost and/or fixed charge rate to override the normal calculations. Transmission and distribution costs are not included. Fuel costs must be input by the user. 2 - Method of solution: The code follows the guidelines of AEC Report NUS-531. A base capital-cost model and a base operating- and maintenance-cost model are selected and adjusted for desired size, location, date, etc. Costs are discounted to the year of first commercial operation and levelized to provide annual cost of electric power generation. 3 - Restrictions on the complexity of the problem: The capital cost models are of doubtful validity outside the 500 to 1500 MW(e) range

  2. Improvement technique of sensitized HAZ by GTAW cladding applied to a BWR power plant

    International Nuclear Information System (INIS)

    Tujimura, Hiroshi; Tamai, Yasumasa; Furukawa, Hideyasu; Kurosawa, Kouichi; Chiba, Isao; Nomura, Keiichi.

    1995-01-01

    A SCC(Stress Corrosion Cracking)-resistant technique, in which the sleeve installed by expansion is melted by GTAW process without filler metal with outside water cooling, was developed. The technique was applied to ICM (In-Core Monitor) housings of a BWR power plant in 1993. The ICM housings of which materials are type 304 Stainless Steels are sensitized with high tensile residual stresses by welding to the RPV (Reactor Pressure Vessel). As the result, ICM housings have potential of SCC initiation. Therefore, the improvement technique resistant to SCC was needed. The technique can improve chemical composition of the housing inside and residual stresses of the housing outside at the same time. Sensitization of the housing inner surface area is eliminated by replacing low-carbon with proper-ferrite microstructure clad. High tensile residual stresses of housing outside surface area is improved into compressive side. Compressive stresses of outside surface are induced by thermal stresses which are caused by inside cladding with outside water cooling. The clad is required to be low-carbon metal with proper ferrite and not to have the new sensitized HAZ (Heat Affected Zone) on the surface by cladding. The effect of the technique was qualified by SCC test, chemical composition check, ferrite content measurement and residual stresses measurement etc. All equipment for remote application were developed and qualified, too. The technique was successfully applied to a BWR plant after sufficient training

  3. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  4. Completion of high-efficiency BWR turbine plant 'Hamaoka unit No. 4'

    International Nuclear Information System (INIS)

    Tsuji, Kunio; Hamaura, Norikazu; Shibashita, Naoaki; Kazama, Seiichi

    1995-01-01

    Accompanying the increase of capacity of nuclear power plants in Japan, the plants having heightened economical efficiency, which are supported by the improvement of thermal efficiency and the reduction of dose, are demanded. Hitachi Ltd. has completed No. 4 turbine unit of 1137 MW output in Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., which is the largest capacity machine in Japanese BWR plants. In this unit, the moisture separator heater, the steam turbine with high efficiency, and the hollow thread film condensate filter which treats the total flow rate of condensate are used as the reheating type BWR plant for the first time in Japan, and the plan of heightened economy and operation was adopted. It was confirmed by the trial for about 10 months that the planned performance was sufficiently satisfied, and the commercial operation was started in September, 1993. The features of the 1137 MW turbine unit are explained. The turbine is of tandem six-flow exhaust condensation type. Diffuser type low pressure turbine exhaust chambers, butterfly type combination intermediate valve are adopted. The stages with the blades having moisture-separating grooves were corrected. The reliability of the shaft system was improved. The adoption of the moisture separator heater and the application of the hollow thread film type condensate filter are explained. (K.I.)

  5. Cooperative control scheme for an HVDC system connected to an isolated BWR nuclear power plant

    International Nuclear Information System (INIS)

    Sakurai, T.; Goto, K.; Kawai, T.; Matori, I.; Nakao, T.; Watanabe, A.

    1983-01-01

    This paper describes a cooperative control system to achieve stable operation of an isolated BWR nuclear plant linked to an HVDC system. In the proposed control system, under normal conditions the power plant is controlled according to the generating power reference and the generator frequency deviation is adjusted by converter power control. Such frequency control is also effective in the case of AC-DC system faults. In addition to the frequency control, an overload control is provided with the HVDC system, where the DC transmission power in the sound poles is increased due to a fault detection signal from the faulty pole. Effects of the above mentioned control systems were studied using digital dynamic programs. The sets of simulation results confirmed that in the case of a DC single pole fault, the plant is able to continue operation without any use of the turbine speed control units even for a restarting failure in the faulty pole. In case of a DC two pole fault, the plant is able to continue operation, being assisted by turbine speed control units when restarting in the faulty poles succeeds. In case of an AC three-line to ground fault near the AC terminal of the converter at the sending or receiving end, the system is able to continue stable operation, being supplemented by the turbine control unit when the faulty section of the AC system is isolated by a main or back-up relaying system

  6. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  7. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

    2001-01-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  8. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  9. HVDC transmission from nuclear power plant

    International Nuclear Information System (INIS)

    Yoshida, Yukio; Takenaka, Kiyoshi; Taniguchi, Haruto; Ueda, Kiyotaka

    1980-01-01

    HVDC transmission directly from a nuclear power plant is expected as one of the bulk power transmission systems from distant power generating area. Successively from the analysis of HVDC transmission from BWR-type nuclear power plant, this report discusses dynamic response characteristics of HVDC transmission (double poles, two circuits) from PWR type nuclear power plant due to dc-line faults (DC-1LG, 2LG) and ac-line faults (3LG) near inverter station. (author)

  10. LWR nuclear power plant component failures

    International Nuclear Information System (INIS)

    Schmidt, W.H.

    1980-10-01

    An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

  11. An application of risk-informed evaluation on MOVs and AOVs for Taiwan BWR-type nuclear power plants

    International Nuclear Information System (INIS)

    Ting, K.; Chen, K.T.; Li, Y.C.; Hwang, S.H.; Chien, F.T.; Kang, J.C.

    2008-01-01

    Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs' and AOVs' safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee

  12. Operator training simulator for nuclear power plant

    International Nuclear Information System (INIS)

    Shiozuka, Hiromi

    1977-01-01

    In nuclear power plants, training of the operators is important. In Japan, presently there are two training centers, one is BWR operation training center at Okuma-cho, Fukushima Prefecture, and another the nuclear power generation training center in Tsuruga City, Fukui Prefecture, where the operators of PWR nuclear power plants are trained. This report describes the BWR operation training center briefly. Operation of a nuclear power plant is divided into three stages of start-up, steady state operation, and shut down. Start-up is divided into the cold-state start-up after the shut down for prolonged period due to periodical inspection or others and the hot-state start-up from stand-by condition after the shut down for a short time. In the cold-state start-up, the correction of reactivity change and the heating-up control to avoid excessive thermal stress to the primary system components are important. The BWR operation training center offers the next three courses, namely beginner's course, retraining course and specific training course. The training period is 12 weeks and the number of trainees is eight/course in the beginner's course. The simulator was manufactured by modeling No. 3 plant of Fukushima First Nuclear Power Station, Tokyo Electric Power Co. The simulator is composed of the mimic central control panel and the digital computer. The software system comprises the monitor to supervise the whole program execution, the logic model simulating the plant interlock system and the dynamic model simulating the plant physical phenomena. (Wakatsuki, Y.)

  13. Computerized operator support system with new man-machine interface for BWR power plants

    International Nuclear Information System (INIS)

    Monta, K.; Naito, N.; Sugawara, M.; Sato, N.; Mori, N.; Tai, I.; Fukumoto, A.; Tsuchida, M.

    1984-01-01

    Improvement of the man-machine interface of nuclear power plants is an important contribution to the further enhancement of operational safety. In addition, recent advances in computer technology seem to offer the greatest opportunity to date for achieving improvement in the man-machine interface. The development of a computerized operator support system for BWRs has been undertaken since 1980 with the support of the Japanese Government. The conceptual design of this system is based on the role of the operators. The main functions are standby system management, disturbance analysis and post-trip operational guidance. The objective of the standby system management is to monitor the standby status of the engineered safety feature during normal operation to assure its proper functioning at the onset of emergency situations. The disturbance analysis system detects disturbances in the plant in their early stages and informs the plant operators about, for example, the cause of the disturbances, the plant status and possible propagations. Consequently, operators can take corrective actions to prevent unnecessary plant shutdown. The objective of the post trip operational guide is to support operators in diagnosis and corrective action after a plant trip. Its functions are to monitor the performance of the engineered safety feature, to identify the plant status and to guide the appropriate corrective action to achieve safe plant shutdown. The information from the computerized operator support system is supplied to operators through a colour CRT operator console. The authors have evaluated the performance of various new man-machine interfacing tools and proposed a new operator console design. A prototype system has been developed and verification/validation is proceeding with a BWR plant simulator. (author)

  14. Availability analysis of United States BWR IV electrical generation plants

    International Nuclear Information System (INIS)

    Renick, D.H.; Li, F.; Todreas, N.E.

    1998-01-01

    Availability, as quantified by power output levels, from all active U.S. BWR IV plants were analyzed over a seven and a half year period to determine the operational characteristics of these plants throughout an operating cycle. The operational data were examined for infant mortality, end of cycle decreased availability, and seasonal availability variations. Scheduled outages were also examined to determine the industry's current approach to planning maintenance outages. The results of this study show that nuclear power plants do suffer significant infant mortality following a refueling outage. And while they do not suffer an end of cycle decrease in availability, a mid-cycle period of decreased availability is evident. This period of decreased availability is due to a combination of increased forced unavailability and seasonally scheduled maintenance and refueling outages. These findings form the start of a rational approach to increasing plant availability. (author)

  15. Development of technologies on innovative-simplified nuclear power plant using high-efficiency steam injectors. (2) Analysis of heat balance of innovative-simplified nuclear power plant

    International Nuclear Information System (INIS)

    Goto, Shoji; Ohmori, Shuichi; Mori, Mitchitsugu

    2004-01-01

    It is possible to established simplified systems and reduced space and equipments using high-efficiency Steam Injector (SI) instead of low-pressure feed water heaters in Nuclear Power Plant (NPP). The SI works as a heat exchanger through direct contact between feedwater from condenser and extracted steam from turbine. It can get a higher pressure than supplied steam pressure, so it can reduce the feedwater pumps. The maintenance and reliability are still higher because SI has no movable parts. This paper describes the analysis of the heat balance and plant efficiency of this Innovative-Simplified NPP with high-efficiency SI. The plant efficiency is compared with the electric power of 1100MWe class original BWR system and the Innovative-Simplified BWR system with SI. The SI model is adapted into the heat balance simulator with a simplified model. The results show plant efficiencies of the Innovated-Simplified BWR system are almost equal to the original BWR one. The present research is one of the projects that are carried out by Tokyo Electric Power Company, Toshiba Corporation, and six Universities in Japan, funded from the Institute of Applied Energy (IAE) of Japan as the national public research-funded program. (author)

  16. Advanced chemistry management system for nuclear power plants

    International Nuclear Information System (INIS)

    Maeda, Katsuji; Kobayashi, Yasuhiro; Nagasawa, Katsumi

    2000-01-01

    Chemistry control in a boiling water reactor (BWR) plant has a close relationship with radiation field buildup, fuel reliability, integrity of plant components and materials, performance of the water treatment systems and radioactive waste generation. Chemistry management in BWR plants has become more important in order to maintain and enhance plant reliability. Adequate chemistry control and management are also essential to establish, maintain, and enhance plant availability. For these reasons, we have developed the advanced chemistry management system for nuclear power plants in order to effectively collect and evaluate a large number of plant operating and chemistry data. (author)

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

    International Nuclear Information System (INIS)

    1992-09-01

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

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

    International Nuclear Information System (INIS)

    1992-09-01

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

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

    International Nuclear Information System (INIS)

    1992-09-01

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

  20. Investigation of valve failure problems in LWR power plants

    International Nuclear Information System (INIS)

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

  1. 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)

  2. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Ito, Noboru; Higuchi, Suminori; Kobayashi, Kensuke; Tobioka, Toshiaki

    1987-03-01

    The PPD (Nuclear Power Plant Data Base) has been under development in JAERI since 1983 as a six-year program to provide useful information for reactor safety regulation and reactor safety research. Information source of the PPD is mainly based on SAR's (Safety Analysis Reports) of 35 nuclear power plants which are operating, under construction or under licensing review in Japan. The report BWR edition consists of several data lists stored in the PPD, concerning safety design such as performances, equipments and installations of 18 BWR plants in Japan. The informations are based on SAR Attachment Chapter 8 ''Nuclear Reactor Facility Safety Design''. (author)

  3. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Horikami, Kunihiko; Kobayashi, Kensuke; Namatame, Ken.

    1993-01-01

    The development of the database called PPD (Nuclear Power Plant Database) has started in 1983 at JAERI as a six-year program to provide useful information for reactor safety regulation and reactor safety research. In 1988 the program has been accomplished, and since then the data in the database has been updating and adding. Information source of the PPD is based on SAR's (Safety Analysis Report) of 47 nuclear power plants which are operating, under construction or under licensing review in Japan. The report, BWR edition, consists of lists of major data stored in the PPD, relating to safety design of 25 BWR plants in Japan. (author)

  4. A nonlinear 3D real-time model for simulation of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.

    1982-02-01

    A nonlinear transient model for BWR nuclear power plants which consists of a 3D-core (subdivided into a number of superboxes, and with parallel flow and subcooled boiling), a top plenum, steam removal and feed water systems and main coolant recirculation pumps is given. The model describes the local core and global plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. The model which forms the basis for the digital code GARLIC-B (Er et al. 82) is aimed to be used on an on-site process computer in parallel to the actual reactor process (or even in predictive mode). Thus, special measures had to be taken into account in order to increase the computational speed and reduce the necessary computer storage. This could be achieved by - separating the neutron and power kinetics from the xenon-iodine dynamics, - treating the neutron kinetics and most of the thermodynamics and hydrodynamics in a pseudostationary way, - developing a special coupling coefficient concept to describe the neutron diffusion, calculating the coupling coefficients from a basic neutron kinetics code, - combining coarse mesh elements into superboxes, taking advantage of the symmetry properties of the core and - applying a sparse matrix technique for solving the resulting algebraic power equation system. (orig.) [de

  5. Reducing BWR O and M costs through on-line performance monitoring

    International Nuclear Information System (INIS)

    Jonas, T.; Gross, R.; Logback, F.; Josyula, R.

    1995-01-01

    Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants

  6. Experiences with monitoring and control of microbiological growth in the standby service water system of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Zisson, P.S.; Whitaker, J.M.; Neilson, H.L.; Mayne, L.L.

    1995-01-01

    In 1989, the Unites States Nuclear Regulatory Commission formally recognized the potential for nuclear accidents resulting from microbiological causes. Such causes range from loss of heat transfer due to microbiological fouling, to loss of system integrity caused by microbiologically influenced corrosion (MIC). As a result of these potential problems, monitoring, mitigation, and control procedures must be developed by all regulated plants. In developing a control and mitigation strategy for the standby service water system of a boiling water reactor (BWR) nuclear power plant, numerous monitoring techniques were employed to evaluate effectiveness. This paper describes the monitoring techniques that were evaluated, and those that ultimately proved to be effective

  7. Investigation of valve failure problems in LWR power plants

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

  8. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  9. Development of technologies on innovative-simplified nuclear power plant using high-efficiency steam injectors (2) analysis of heat balance of innovative-simplified nuclear power plant

    International Nuclear Information System (INIS)

    Goto, S.; Ohmori, S.; Mori, M.

    2005-01-01

    It is possible to establish simplified system with reduced space and total equipment weight using high-efficiency Steam Injector (SI) instead of low-pressure feedwater heaters in Nuclear Power Plant (NPP)(1)-(6). The SI works as a heat exchanger through direct contact between feedwater from the condensers and extracted steam from the turbines. It can get a higher pressure than supplied steam pressure, so it can reduce the feedwater pumps. The maintenance and reliability are still higher because SI has no movable parts. This paper describes the analysis of the heat balance and plant efficiency of this Innovative- Simplified NPP with high-efficiency SI. The plant efficiency is compared with the electric power of 1100MWe-class BWR system and the Innovative- Simplified BWR system with SI. The SI model is adapted into the heat balance simulator with a simplified model. The results show plant efficiencies of the Innovated-Simplified BWR system are almost equal to the original BWR one. The present research is one of the projects that are carried out by Tokyo Electric Power Company, Toshiba Corporation, and six Universities in Japan, funded from the Institute of Applied Energy (IAE) of Japan as the national public research-funded program. (authors)

  10. Seismic risk assessment of a BWR: status report

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

    1985-02-01

    The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

  11. ASTM standards associated with PWR and BWR power plant licensing, operation and surveillance

    International Nuclear Information System (INIS)

    McElroy, W.N.; McElroy, R.J.; Gold, R.; Lippincott, E.P.; Lowe, A.L. Jr.

    1994-01-01

    This paper considers ASTM Standards that are available, under revision, and are being considered in support of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) Nuclear Power Plant (NPP) licensing, regulation, operation, surveillance and life attainment. The current activities of ASTM Committee E10 and its Subcommittees E10.02 and current activities of ASTM Committee E10 and its Subcommittees E10.02 and E10.05 and their Task Groups (TG) are described. A very important aspect of these efforts is the preparation, revision, and balloting of standards identified in the ASTM E706 Standard on Master Matrix for Light Water Reactor (LWR) Pressure Vessel (PV) Surveillance Standards. The current version (E706-87) of the Master Matrix identifies 21 ASTM LWR physics-dosimetry-metallurgy standards for Reactor Pressure Vessel (RPV) and Support Structure (SS) surveillance programs, whereas, for the next revision 34 standards are identified. The need for national and international coordination of Standards Technology Development, Transfer and Training (STDTT) is considered in this and other Symposium papers that address specific standards related physics-dosimetry-metallurgy issues. 69 refs

  12. HVDC transmission from nuclear power plant

    International Nuclear Information System (INIS)

    Yoshida, Yukio; Takenaka, Kiyoshi; Ichikawa, Takemi; Ueda, Kiyotaka; Machida, Takehiko

    1979-01-01

    The HVDC transmission directly from nuclear power plants is one of the patterns of long distance and large capacity HVDC transmission systems. In this report, the double pole, two-circuit HVDC transmission from a BWR nuclear power plant is considered, and the dynamic response characteristics due to the faults in dc line and ac line of inverter side are analyzed, to clarify the dynamic characteristics of the BWR nuclear power plant and dc system due to system faults and the effects of dc power control to prevent reactor scram. (1) In the instantaneous earthing fault of one dc line, the reactor is not scrammed by start-up within 0.8 sec. (2) When the earthing fault continues, power transmission drops to 75% by suspending the faulty pole, and the reactor is scrammed. (3) In the instantaneous ground fault of 2 dc lines, the reactor is not scrammed if the faulty dc lines are started up within 0.4 sec. (4) In the existing control of dc lines, the reactor is scrammed when the ac voltage at an ac-dc connection point largely drops due to ac failure. (J.P.N.)

  13. Technical description and evaluation of BWR hybrid power shape monitoring system. Final report

    International Nuclear Information System (INIS)

    Frogner, B.; Ipaktchi, A.; Yang, C.; Grow, R.; Ho, C.; Kiguchi, T.

    1982-03-01

    This report discusses the method for monitoring BWR cores that has been implemented in the Power Shape Monitoring System (PSMS). The approach has been benchmarked to TIP data from three plants and five fuel cycles and the accuracy of the calculations has been evaluated by using gamma scan data from two plants. A coupled neutronics/thermal-hydraulic nodal code (NODE-B/THERM-B) is used in the PSMS. It has been demonstrated that adaptation of this code to partially fit the TIP readings followed by a statistical characterization of the remaining errors results in better accuracy and improved sensitivity for anomaly detection compared to an approach that is entirely dependent upon the detector readings. The computed power distribution has a one-sigma uncertainty of 6% for the nodal power and 4% for the bundle power. This is significantly better than the plant process computers that actually were used for monitoring those two plants where comparisons were made

  14. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor

    International Nuclear Information System (INIS)

    Escorcia O, D.; Salazar S, E.

    2016-09-01

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  15. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  16. 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

  17. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  18. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  19. Innovative-Simplified Nuclear Power Plant Efficiency Evaluation with High-Efficiency Steam Injector System

    International Nuclear Information System (INIS)

    Shoji, Goto; Shuichi, Ohmori; Michitsugu, Mori

    2006-01-01

    It is possible to establish simplified system with reduced space and total equipment weight using high-efficiency Steam Injectors (SI) instead of low-pressure feedwater heaters in Nuclear Power Plant (NPP). The SI works as a heat exchanger through direct contact between feedwater from condensers and extracted steam from turbines. It can get higher pressure than supplied steam pressure. The maintenance and reliability are still higher than the feedwater ones because SI has no movable parts. This paper describes the analysis of the heat balance, plant efficiency and the operation of this Innovative-Simplified NPP with high-efficiency SI. The plant efficiency and operation are compared with the electric power of 1100 MWe-class BWR system and the Innovative-Simplified BWR system with SI. The SI model is adapted into the heat balance simulator with a simplified model. The results show that plant efficiencies of the Innovated-Simplified BWR system are almost equal to original BWR ones. The present research is one of the projects that are carried out by Tokyo Electric Power Company, Toshiba Corporation, and six Universities in Japan, funded from the Institute of Applied Energy (IAE) of Japan as the national public research-funded program. (authors)

  20. Power plant design: ESBWR - the latest passive BWR

    International Nuclear Information System (INIS)

    Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

    1997-01-01

    When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

  1. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  2. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

    Smith, S.K.; Lehnert, D.F.; Locke, R.K.

    1991-01-01

    Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

  3. Automatic refueling platform and CRD remote handling device for BWR plant

    International Nuclear Information System (INIS)

    Kato, Hiroaki; Takagi, Kaoru

    1978-01-01

    In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

  4. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    1999-05-01

    This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

  5. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  6. Construction techniques and management methods for BWR plants

    International Nuclear Information System (INIS)

    Shimizu, Yohji; Tateishi, Mizuo; Hayashi, Yoshishige

    1989-01-01

    Toshiba is constantly striving for safer and more efficient plant construction to realize high-quality BWR plants within a short construction period. To achieve these aims, Toshiba has developed and improved a large number of construction techniques and construction management methods. In the area of installation, various techniques have been applied such as the modularization of piping and equipment, shop installation of reactor internals, etc. Further, installation management has been upgraded by the use of pre-installation review programs, the development of installation control systems, etc. For commissioning, improvements in commissioning management have been achieved through the use of computer systems, and testing methods have also been upgraded by the development of computer systems for the recording and analysis of test data and the automatic adjustment of controllers in the main control system of the BWR. This paper outlines these construction techniques and management methods. (author)

  7. Power generation by nuclear power plants

    International Nuclear Information System (INIS)

    Bacher, P.

    2004-01-01

    Nuclear power plays an important role in the world, European (33%) and French (75%) power generation. This article aims at presenting in a synthetic way the main reactor types with their respective advantages with respect to the objectives foreseen (power generation, resources valorization, waste management). It makes a fast review of 50 years of nuclear development, thanks to which the nuclear industry has become one of the safest and less environmentally harmful industry which allows to produce low cost electricity: 1 - simplified description of a nuclear power generation plant: nuclear reactor, heat transfer system, power generation system, interface with the power distribution grid; 2 - first historical developments of nuclear power; 3 - industrial development and experience feedback (1965-1995): water reactors (PWR, BWR, Candu), RBMK, fast neutron reactors, high temperature demonstration reactors, costs of industrial reactors; 4 - service life of nuclear power plants and replacement: technical, regulatory and economical lifetime, problems linked with the replacement; 5 - conclusion. (J.S.)

  8. Dose rate reduction method for NMCA applied BWR plants

    International Nuclear Information System (INIS)

    Nagase, Makoto; Aizawa, Motohiro; Ito, Tsuyoshi; Hosokawa, Hideyuki; Varela, Juan; Caine, Thomas

    2012-09-01

    BRAC (BWR Radiation Assessment and Control) dose rate is used as an indicator of the incorporation of activated corrosion by products into BWR recirculation piping, which is known to be a significant contributor to dose rate received by workers during refueling outages. In order to reduce radiation exposure of the workers during the outage, it is desirable to keep BRAC dose rates as low as possible. After HWC was adopted to reduce IGSCC, a BRAC dose rate increase was observed in many plants. As a countermeasure to these rapid dose rate increases under HWC conditions, Zn injection was widely adopted in United States and Europe resulting in a reduction of BRAC dose rates. However, BRAC dose rates in several plants remain high, prompting the industry to continue to investigate methods to achieve further reductions. In recent years a large portion of the BWR fleet has adopted NMCA (NobleChem TM ) to enhance the hydrogen injection effect to suppress SCC. After NMCA, especially OLNC (On-Line NobleChem TM ), BRAC dose rates were observed to decrease. In some OLNC applied BWR plants this reduction was observed year after year to reach a new reduced equilibrium level. This dose rate reduction trends suggest the potential dose reduction might be obtained by the combination of Pt and Zn injection. So, laboratory experiments and in-plant tests were carried out to evaluate the effect of Pt and Zn on Co-60 deposition behaviour. Firstly, laboratory experiments were conducted to study the effect of noble metal deposition on Co deposition on stainless steel surfaces. Polished type 316 stainless steel coupons were prepared and some of them were OLNC treated in the test loop before the Co deposition test. Water chemistry conditions to simulate HWC were as follows: Dissolved oxygen, hydrogen and hydrogen peroxide were below 5 ppb, 100 ppb and 0 ppb (no addition), respectively. Zn was injected to target a concentration of 5 ppb. The test was conducted up to 1500 hours at 553 K. Test

  9. Condensate treatment and oxygen control in power plants

    International Nuclear Information System (INIS)

    Sakai, Toshiaki; Iida, Kei; Ohashi, Shinichi.

    1997-01-01

    In thermal and nuclear power stations, the steam that operated turbines is cooled and condensed with condensers. The condensate is heated again with boilers, nuclear reactors or steam generators, but if corrosion products or impurities are contained in the condensate, corrosion and scale formation occur in boilers and others. The filtration facility and the desalting facility for condensate are installed to remove impurities, but water quality control is different in thermal, BWR and PWR plants, therefore, the treatment facilities corresponding to respective condensates have been adopted. In order to reduce the amount of clud generation, the treatment of injecting a small quantity of oxygen into condensate has been adopted. In thermal power plants, all volatile treatment is carried out, in which corrosion is prevented by the addition of ammonia and hydrazine to boiler feedwater. The condensate filters of various types and the NH 4 type condensate desalter for thermal power plants are described. In BWR power plants, steam is generated in nuclear reactors, therefore, the addition of chemicals into water is never carried out, and high purity neutral water is used. In PWR power plants, the addition of chemicals to water is done in the primary system, and AVT is adopted in the secondary system. Also the condensate treatment facilities are different for both reactors. (K.I.)

  10. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

  11. Turbine protecting device in a BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kasuga, Hajime; Oka, Yoko.

    1984-01-01

    Purpose: To prevent highly humid steams from flowing into the turbine upon abnormal reduction in the reactor water level in order to ensure the turbine soundness, as well as in order to trip the turbine with no undesired effect on the reactor. Constitution: A protection device comprising a judging device and a timer are disposed in a BWR type reactor, in order to control a water level signal from a reactor water level gage. If the reactor water level is reduced during rated power operation, steams are kept to be generated due to decay heat although reactor is scramed. When a signal from the reactor water level detector is inputted to the protection device, a trip signal is outputted by way of a judging device after 15 second by means of the timer, when the main steam check valve is closed to trip the turbine. With this delay of time, abrupt increase in the pressure of the reactor due to sudden shutdown can be prevented. (Nakamoto, H)

  12. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  13. Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

    1996-01-01

    A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

  14. Radiochemistry in nuclear power plants

    International Nuclear Information System (INIS)

    Schwarz, W.

    2007-01-01

    Radiochemistry is employed in nuclear power plants not as an end in itself but, among other things, as a main prerequisite of optimum radiation protection. Radiochemical monitoring of various loops provides important information about sources of radioactivity, activity distribution in the plant and its changes. In the light of these analytical findings, plant crews are able to take measures having a positive effect on radiation levels in the plant. The example of a BWR plant is used to show, among other things, how radiochemical analyses helped to reduce radiation levels in a plant and, as a consequence, to decrease clearly radiation exposure of the personnel despite higher workloads. (orig.)

  15. Nuclear power plant

    International Nuclear Information System (INIS)

    Aisaka, Tatsuyoshi; Kamahara, Hisato; Yanagisawa, Ko.

    1982-01-01

    Purpose: To prevent corrosion stress cracks in structural materials in a BWR type nuclear power plant by decreasing the oxygen concentration in the reactor coolants. Constitution: A hydrogen injector is connected between the condensator and a condensate clean up system of a nuclear power plant. The injector is incorporated with hydrogenated compounds formed from metal hydrides, for example, of alloys such as lanthanum-nickel alloy, iron titanium alloy, vanadium, palladium, magnesium-copper alloy, magnesium-nickel alloy and the like. Even if the pressure of hydrogen obtained from a hydrogen bomb or by way of water electrolysis is changed, the hydrogen can always be injected into a reactor coolant at a pressure equal to the equilibrium dissociation pressure for metal hydride by introducing the hydrogen into the hydrogen injector. (Seki, T.)

  16. On-line prediction of BWR transients in support of plant operation and safety analyses

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

    1983-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feed-water train. Point kinetics incorporate reactivity feedback for void fraction, for fuel temperature, and for coolant temperature. Control systems and trip logic are simulated for the nuclear steam supply system

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

    International Nuclear Information System (INIS)

    1995-04-01

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

  18. General model for Pc-based simulation of PWR and BWR plant components

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

    1995-10-01

    In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

  19. Dura Seal recommendations for nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Reactor systems (BWR, PWR and Candu) are briefly reviewed with reference to the pumping services encountered in each system, to indicate the conditions imposed on mechanical seals for nuclear power plant liquid handling equipment. A description of the Dura Seals used in each service is included. (U.K.)

  20. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  1. GARLIC-B. A digital code for real-time calculation of the transient behaviour of nodal and global core and plant parameters of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.; Hoeld, A.; Lupas, O.

    1982-04-01

    A program description of the code GARLIC-B is given. The code is based on a nonlinear transient model for BWR nuclear power plants which consist of a 3D-core, a top plenum, steam removal and feed water systems and a downcomer with main coolant recirculation pumps. The core is subdivided into a number of superboxes and flow channels with different coolant mass flow rates. Subcooled boiling within these channels has an important reactivity feed back effect and has to be taken also into account. The code computes the local and global core and plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. (orig./GL) [de

  2. Flex concept for US-A BWR extended loss of AC power events

    International Nuclear Information System (INIS)

    Powers, J.; Aoyagi, Y.; Kataoka, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  3. Flex concept for US-A BWR extended loss of AC power events

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  4. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  5. Water chemistry experience following an extensive power up-rate in Oskarshamn 3 BWR

    International Nuclear Information System (INIS)

    Wegemar, Boerje; Nilsson, Jimmy; Lejon Johan; Bergfors, Asa; Arnberg, Bo

    2012-09-01

    The Swedish Oskarshamn 3 BWR plant, operated by OKG, was first connected to the grid in 1985. The plant has been power up-rated in two steps; from the original design, 3020 MWth, to 3300 MWth (109%, 1989) and recently to 3900 MWth (129%, 2009). Westinghouse Electric Sweden AB (former ASEA-Atom, OEM of the plant) was rewarded a major contract in the recently implemented up-rating project, the PULS project. The PULS project is quite unique since no operating experience has to date been reported from a similar major power up-rate in a BWR plant. Water chemistry experience from the first period of operation following the implementation of the PULS project is reported and discussed in the paper. Reported chemistry and radiochemistry measurements in feedwater (FW) and reactor water (RW) include corrosion products, activated corrosion products, dissolved oxygen and impurities like chloride, sulfate etc. Furthermore, a comparison of water quality prior to implementation of the PULS project is included. Several process systems have been modified, one of them being the condensate cleanup system (CCU), a Pre-coat filter system. The design criteria for the CCU system include the filter run-lengths, pressure drop before back-washing and requirements on water chemistry quality. The paper describes in some detail the CCU system modifications being implemented in order to fulfil the design criterion. CCU cleanup efficiency, operating temperature and influence of hydrogen peroxide on the CCU resin are all important issues being covered in the paper. As for the latter, it is well known that oxygen and hydrogen peroxide (from radiolysis in the core region) might cause partial deterioration of CCU standard cation resin resulting in increased RW sulfate concentrations. This aspect is covered in the paper as well. The reactor water cleanup system (RWCU) in Oskarshamn 3 consists of deep bed ion exchange filters (mixed bed filter). The purpose of RWCU is to maintain a low level of

  6. Experience on a BWR plant diagnosis system

    International Nuclear Information System (INIS)

    Tanabe, A.; Kawai, K.; Hashimoto, Y.

    1981-01-01

    It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

  7. Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly

    International Nuclear Information System (INIS)

    Mitsutake, T.; Terasaka, H.; Yoshimura, K.; Oishi, M.; Inoue, A.; Akiyama, M.

    1990-01-01

    Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8x8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F 'multi-fluid' computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, though comparisons, between the prediction and the obtained test data. (orig.)

  8. BWR shutdown analyzer using artificial intelligence (AI) techniques

    International Nuclear Information System (INIS)

    Cain, D.G.

    1986-01-01

    A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

  9. New design procedure development of future reactor critical power estimation. (1) Practical design-by-analysis method for BWR critical power design correlation

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Mitsutake, Toru

    2007-01-01

    For present BWR fuels, the full mock-up thermal-hydraulic test, such as the critical power measurement test, pressure drop measurement test and so on, has been needed. However, the full mock-up test required the high costs and large-scale test facility. At present, there are only a few test facilities to perform the full mock-up thermal-hydraulic test in the world. Moreover, for future BWR, the bundle size tends to be larger, because of reducing the plant construction costs and minimizing the routine check period. For instance, AB1600, improved ABWR, was proposed from Toshiba, whose bundle size was 1.2 times larger than the conventional BWR fuel size. It is too expensive and far from realistic to perform the full mock-up thermal-hydraulic test for such a large size fuel bundle. The new design procedure is required to realize the large scale bundle design development, especially for the future reactor. Therefore, the new design procedure, Practical Design-by-Analysis (PDBA) method, has been developed. This new procedure consists of the partial mock-up test and numerical analysis. At present, the subchannel analysis method based on three-fluid two-phase flow model only is a realistic choice. Firstly, the partial mock-up test is performed, for instance, the 1/4 partial mock-up bundle. Then, the first-step critical power correlation coefficients are evaluated with the measured data. The input data, such as the spacer effect model coefficient, on the subchannel analysis are also estimated with the data. Next, the radial power effect on the critical power of the full-bundle size was estimated with the subchannel analysis. Finally, the critical power correlation is modified by the subchannel analysis results. In the present study, the critical power correlation of the conventional 8x8 BWR fuel was developed with the PDBA method by 4x4 partial mock-up tests and the subchannel analysis code. The accuracy of the estimated critical power was 3.8%. The several themes remain to

  10. BWR plant advanced central control panel PODIA

    International Nuclear Information System (INIS)

    Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

    1983-01-01

    BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

  11. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  12. A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

    International Nuclear Information System (INIS)

    Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

    2017-01-01

    This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

  13. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  14. Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

    International Nuclear Information System (INIS)

    2000-10-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

  15. Recent control and instrumentation systems for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Fujii, Hiroaki; Higashikawa, Yuichi; Sato, Hideyuki

    1990-01-01

    For the needs of the more stable operation of nuclear power stations, the upgrading of the measurement and control system for BWRs has been promoted by positively introducing remarkably advancing electronic technology. Further, it is aimed at to construct the synthetic digitized measurement and control system for nuclear power stations to heighten the operation reliability in ABWRs. As the first step of the development in the synthetic digitization, the monitoring and control system for radioactive waste treatment was put in practical use for No.5 plant of Kashiwazaki, Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. Hitachi Ltd. has promoted the development and the application to actual plants of the measurement and control system for BWRs, in which digital control technology, optical information transmission technology and the operation-supporting technology using a computer were utilized. Hereafter, it is intended to expand the application of digital measurement and control aiming at improving the reliability, operation performance and maintainability. The nuclear power plant control complex with advanced man-machine interface-90 (NUCAMM-90) was developed, and its application to actual plants is planned. (K.I.)

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

    International Nuclear Information System (INIS)

    1995-04-01

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

  17. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    International Nuclear Information System (INIS)

    Aounallah, Y.

    2001-01-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  18. Development of vendor independent safety analysis capability for nuclear power plants in Taiwan

    International Nuclear Information System (INIS)

    Tang, J.-R.

    2001-01-01

    The Institute of Nuclear Energy Research (INER) and the Taiwan Power Company (TPC) have long-term cooperation to develop vendor independent safety analysis capability to provide support to nuclear power plants in Taiwan in many aspects. This paper presents some applications of this analysis capability, introduces the analysis methodology, and discusses the significance of vendor independent analysis capability now and future. The applications include a safety analysis of core shroud crack for Chinshan BWR/4 Unit 2, a parallel reload safety analysis of the first 18-month extended fuel cycle for Kuosheng BWR/6 Unit 2 Cycle 13, an analysis to support Technical Specification change for Maanshan three-loop PWR, and a design analysis to support the review of Preliminary Safety Analysis Report of Lungmen ABWR. In addition, some recent applications such as an analysis to support the review of BWR fuel bid for Chinshan and Kuosheng demonstrates the needs of further development of the analysis capability to support nuclear power plants in the 21 st century. (authors)

  19. Application of process computers and colour CRT displays in the plant control room of a BWR

    International Nuclear Information System (INIS)

    Itoh, M.; Hayakawa, H.; Kawahara, H.; Neda, T.; Wakabayashi, Y.

    1983-01-01

    The recent application of a CRT display system in an 1100-MW(e) BWR plant control room and the design features of a new control room whose installation is planned for the next generation are discussed. As reactor unit capacity and the need for plant safety and reliability continue to increase, instrumentation and control equipment is growing in number and complexity. In consequence, control and supervision of plant operations require improvement. Thus, because of recent progress in the field of process computers and display equipment (colour CRTs), efficient improvements of the control room are under way in the Japanese BWR plant. In the recently constructed BWR plant (1100 MW(e)), five CRTs on the bench board and two process computers were additionally installed in the control room during the construction stage to improve plant control and supervisory functions by implementing the lessons learned from the Three Mile Island incident. The major functions of the new computers and display systems are to show integrated graphic displays of the plant status, to monitor the standby condition of the safety system, to show the condition of the integrated alarm system, etc. In practice, in the actual plant, this newly installed system performs well. On the basis of the experience gained in these activities, a new computerized control and monitoring system is now being designed for subsequent domestic BWR plants. This advanced system will incorporate not only the functions already mentioned, but also a surveillance guide system and plant automation. For future plants, a diagnostic system and an instructional system that can analyse a disturbance and give operational guidance to the plant operator are being developed in a government-sponsored programme. (author)

  20. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Cruz G, M.; Amador C, C.

    2013-10-01

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  1. Availability of thermal power plants 1981-1990

    International Nuclear Information System (INIS)

    Nitsch, D.; Schmitz, H.

    1991-01-01

    The present volume covers the period of 1981 to 1990 and contains availability data of power plants in Germany and abroad. Data are presented on fossil-fuelled units, units with a combined gas/steam cycle, nuclear power plants and gas turbines. The fossil-fuelled units are broken down by unit size, years of operation, fuel, type of combustion (dry, melt) and type (mono, duo units, subcritical and supercritical systems). Nuclear power stations are arranged by type of reactor (PWR, BWR), unit size and years of operation. Combined cycle power plants are listed separately due to their different technical concepts. Apart from availability and utilisation values of gas turbines there are data on reliability and the number of successful and unsuccessful starts. In general the data are first given for all plants and then for the German plants in particular. Performance values are gross values measured at generator terminals and, as the number of plants, they are end-of-the-year figures [de

  2. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  3. Development of Halogen-free flame-retardant cable for nuclear power plant

    International Nuclear Information System (INIS)

    Ishii, Nobuhisa; Morii, Akira; Fujimura, Shunichi

    1992-01-01

    Conventional flame-retardant cables release a large volume of corrosive and toxic gases as well as smoke while combusted. Cables covered with halogen-free flame-retardant material, containing no halogen in it, have been developed to reduce generation of such gases and smoke, and have already been used in telecommunication service, subway and shipboard applications. However, for cables for nuclear power plant, covering materials should also have radiation resistance and other properties, including long-term physical stability. We have developed halogen-free flame-retardant cables for BWR nuclear power plant with sufficient flame retardancy radiation resistance and environmental resistance including steam-exposure resistance all of which are in accordance with Japanese specifications for BWR nuclear cables and have such characteristics as low corrosiveness, low toxicity and low smoke emission. (author)

  4. Facility of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsuji

    1998-01-01

    A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

  5. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

    2008-07-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  6. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

    2008-01-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  7. Millstone nuclear power plant emergency system assessment

    International Nuclear Information System (INIS)

    Akhmad Khusyairi

    2011-01-01

    U.S.NRC determined an obligation to build a nuclear power plant emergency response organization for both on-site and off-site. Millstone Nuclear Power Plants have 3 nuclear reactors and 2 of 3 still in commercial operation. Reactor unit 1, BWR type has been permanently shut down in 1998, while the two others, units 2 and 3 obtain the extended operating license respectively until 2035 and 2045. As a nuclear installation has the high potential radiological impact, Millstone nuclear power plant emergency response organization must establish both on-site or off-site. Emergency response organization that is formed must involve several state agencies, both state agencies and municipality. They have specific duties and functions in a state of emergency, so that protective measures can be undertaken in accordance with the community that has been planned. Meanwhile, NRC conduct their own independent assessment of nuclear power plant emergencies. (author)

  8. Sophistication of operator training using BWR plant simulator

    International Nuclear Information System (INIS)

    Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

    1986-01-01

    In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

  9. Valuation of power oscillations in a BWR after control rod banks withdrawal events

    International Nuclear Information System (INIS)

    Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

    2009-01-01

    The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

  10. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  11. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  12. An overview of the BWR ECCS strainer blockage issues

    International Nuclear Information System (INIS)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

    1996-01-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

  13. 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

  14. Comparison of the corrosion potential for stainless steel measured in-plant and in laboratory during BWR normal water chemistry conditions

    International Nuclear Information System (INIS)

    Molander, A.; Pein, K.; Tarkpea, P.; Takagi, Junichi; Karlberg, G.; Gott, K.

    1998-01-01

    To obtain reliable crack growth rate date for stainless steel in BWR environments careful laboratory simulation of the environmental conditions is necessary. In the plant the BWR normal water chemistry environment contains hydrogen peroxide, oxygen and hydrogen. However, in crack growth rate experiments in laboratories, the environment is normally simulated by adding 200 ppb oxygen to the high temperature water. Thus, as hydrogen peroxide is a more powerful oxidant than oxygen, it is to be expected that a lower corrosion potential will be measured in the laboratory than in the plant. To resolve this issue this work has been performed. In-plant and laboratory measurements have often been performed with somewhat different equipment, due to the special requirements concerning in-plant measurements. In this work such differences have been avoided and two identical sets of equipment for electrochemical measurements were built and used for measurements in-plant in a Swedish BWR and in high purity water in the laboratory. The host plant was Barsebaeck 1. Corrosion potential monitoring in-plant was performed under both NWC (Normal Water Chemistry) and HWC (Hydrogen Water Chemistry) conditions. This paper is, however, focused on NWC conditions. This is due to the fact, that the total crack growth obtained during a reactor cycle, can be determined by NWC conditions, even for plants running with HWC due to periodic stops in the hydrogen addition for turbine inspections or failure of the dosage or hydrogen production equipment. Thus, crack growth data for NWC is of great importance both for BWRs operating with HWC and NWC. Measurements in-plant and in the laboratory were performed during additions of oxygen and hydrogen peroxide to the autoclave systems. The corrosion potentials were compared for various conditions in the autoclaves, as well as versus in-plant in-pipe corrosion potentials. (J.P.N.)

  15. Risk evaluation of the alternate-3A modification to the ATWS prevention/mitigation system in a BWR-4, MARK-II power plant

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Karol, R.; Shiu, K.

    1983-01-01

    The authors present a risk evaluation of the ATWS Alternate 3A modification proposed by NRC staff in NUREG-0460 to the ATWS prevention/mitigation system in a BWR nuclear power plant. The evaluation is done relative to three risk indices: the frequency of core damage, the expected early fatalities, and the expected latent fatalities. The ATWS prevention tree includes: the mechanical subsystem of the reactor protection system, the electrical subsystem of the reactor protection system, the recirculation pump trip and the Alternate Rod Insertion System. The mitigation tree includes: standby liquid control system, opening of the relief valves, reclosing the relief valves, failure of coolant injection, inadvertent actuation of the automatic depressurization system, inadvertent operation of high-pressure injection system and containment heat removal

  16. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  17. A real-time expert system for nuclear power plant failure diagnosis and operational guide

    International Nuclear Information System (INIS)

    Naito, N.; Sakuma, A.; Shigeno, K.; Mori, N.

    1987-01-01

    A real-time expert system (DIAREX) has been developed to diagnose plant failure and to offer a corrective operational guide for boiling water reactor (BWR) power plants. The failure diagnosis model used in DIAREX was systematically developed, based mainly on deep knowledge, to cover heuristics. Complex paradigms for knowledge representation were adopted, i.e., the process representation language and the failure propagation tree. The system is composed of a knowledge base, knowledge base editor, preprocessor, diagnosis processor, and display processor. The DIAREX simulation test has been carried out for many transient scenarios, including multiple failures, using a real-time full-scope simulator modeled after the 1100-MW(electric) BWR power plant. Test results showed that DIAREX was capable of diagnosing a plant failure quickly and of providing a corrective operational guide with a response time fast enough to offer valuable information to plant operators

  18. Study on optimization of normal plant outage work plan for nuclear power plants

    International Nuclear Information System (INIS)

    Aoki, Takayuki; Kodama, Noriko; Takase, Kentaro; Miya, Kenzo

    2011-01-01

    This paper discusses maintenance optimization in maintenance implementation stage following maintenance planning stage in nuclear power plants and proposes a methodology to get an optimum maintenance work plan. As a result of consideration, the followings were obtained. (1) The quantitative evaluation methodology for optimizing maintenance work plan in nuclear power plants was developed. (2) Utilizing the above methodology, a simulation analysis of maintenance work planning for BWR's PLR and RHR systems in a normal plant outage was performed. Maintenance cost calculation in several cases was carried out on the condition of smoothening man loading over the plant outage schedule as much as possible. (3) As a result of the simulation, the economical work plans having a flat man loading over the plant outage schedule were obtained. (author)

  19. Feasibility study on development of plate-type heat exchanger for BWR plants

    International Nuclear Information System (INIS)

    Ohyama, Nobuhiro; Suda, Kenichi; Ogata, Hiroshi; Matsuda, Shinichi; Nagasaka, Kazuhiro; Fujii, Toshi; Nozawa, Toshiya; Ishihama, Kiyoshi; Higuchi, Tomokazu

    2004-01-01

    In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

  20. Analysis of heat balance on innovative-simplified nuclear power plant using multi-stage steam injectors

    International Nuclear Information System (INIS)

    Goto, Shoji; Ohmori, Shuichi; Mori, Michitsugu

    2006-01-01

    The total space and weight of the feedwater heaters in a nuclear power plant (NPP) can be reduced by replacing low-pressure feedwater heaters with high-efficiency steam injectors (SIs). The SI works as a direct heat exchanger between feedwater from condensers and steam extracted from turbines. It can attain pressures higher than the supplied steam pressure. The maintenance cost is lower than that of the current feedwater heater because of its simplified system without movable parts. In this paper, we explain the observed mechanisms of the SI experimentally and the analysis of the computational fluid dynamics (CFD). We then describe mainly the analysis of the heat balance and plant efficiency of the innovative-simplified NPP, which adapted to the boiling water reactor (BWR) with the high-efficiency SI. The plant efficiencies of this innovative-simplified BWR with SI are compared with those of a 1 100 MWe-class BWR. The SI model is adopted in the heat balance simulator as a simplified model. The results show that the plant efficiencies of the innovate-simplified BWR with SI are almost equal to those of the original BWR. They show that the plant efficiency would be slightly higher if the low-pressure steam, which is extracted from the low-pressure turbine, is used because the first-stage of the SI uses very low pressure. (author)

  1. Rock siting of nuclear power plants from a reactor safety standpoint. Status report October 1974

    International Nuclear Information System (INIS)

    1975-01-01

    The aim of this study is to clearify the advantages and disadvantages of an underground nuclear power plant from a reactor safety point of view, compared to a plant above ground. Principles for the technical design of a rock sited BWR nuclear power plant is presented. Also questions of sabotage and closing down the plant at the end of the operational period are treated. (K.K.)

  2. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  3. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  4. Characteristic evaluations of BWR uprate method based on heat balance shift concept

    International Nuclear Information System (INIS)

    Kitou, Kazuaki; Aoyama, Motoo; Shiina, Kouji; Sasaki, Hiroshi; Yoshikawa, Kazuhiro

    2007-01-01

    Reactor power uprate of nuclear power plants is an efficient plant operating method. Most BWR plants need the exchange of high pressure turbines when plant thermal power increases over 5% because main steam flow rate exceeds the limitation of inlet steam flow rate of a high pressure turbine. Therefore, the new power uprate method named heat balance shift power uprate method has been developed. This method decreases feedwater temperature with increasing plant thermal power not to increase main steam flower rate. This study clarified that the heat balance shift method could increase plant electric power up to 2.8% compared with conventional power uprate method without the exchange of a high pressure turbine. (author)

  5. Power ramping/cycling experience and operational recommendations in KWU power plants

    International Nuclear Information System (INIS)

    Jan, R. von; Wunderlich, F.; Holzer, R.

    1980-01-01

    The power cycling and ramping experience of KWU is based on experiments in test and commercial reactors, and on evaluation of plant operation (PHWR, PWR and BWR). Power cycling of fuel rods have never lead to PCI failures. In ramping experiments, for fast ramps PCI failure thresholds of 480/420 W/cm are obtained at 12/23 GWd/t(U) burn-up for pressurized PWR fuel. No failures occurred during limited exceedance of the threshold with reduced ramp rate. Operational recommendations used by KWU are derived from experiments and plant experience. The effects of ramping considerations on plant operation is discussed. No rate restrictions are required for start-ups during an operating cycle or load follow operation within set limits for the distortion of the local power distribution. In a few situations, e.g. start-up after refueling, ramp rates of 1 to 5 %/h are recommended depending on plant and fuel design

  6. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

    2010-01-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  7. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

    2010-07-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  8. Design optimization for fuel reloading in Laguna Verde nuclear power plant

    International Nuclear Information System (INIS)

    Cortes Campos, C.C.; Montes Tadeo, J.L.

    1991-01-01

    Procedure followed to perform the design optimation in fuel reloading is described in general words and also is shown an example in which such procedure was uses for analysis of BWR type reactor in unit 1 of Laguna Verde nuclear power plant (Author)

  9. Relative radiation hazards of coal based and nuclear power plants

    International Nuclear Information System (INIS)

    Mishra, U.C.

    1983-04-01

    Coal, like most materials found in nature, contains trace quantities of naturally occurring radionuclides. However, low concentrations may become important if large quantities of coal are burnt in thermal power plants. Therefore a study was performed to determine the radioactivity in coal, in fly-ash and slag and assess the importance of radioactive emissions from thermal power plants. The results were compared to the radiological impact of nuclear power stations. Based on these data, theoretical estimates for the population living within 80km from power stations indicate that the collective dose commitments of coal-fired plants are one order of magnitude higher than those for BWR-type nuclear plants. Measurements taken in the vicinity of coal-fired plants were comparable to those for nuclear plants, i.e. within the range of variation of natural background radiation in India

  10. 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

  11. Power conditioning devices in nuclear power plants

    International Nuclear Information System (INIS)

    Shida, Toichi.

    1979-01-01

    Purpose: To automatically prevent the liquid level from lowering in a reactor even if a feedwater adjusting valve is locked in a bwr type nuclear power plant. Constitution: Where a feedwater adjusting valve should be locked, and if the mismatching degree between the main steam flow rate and the feedwater flow rate exceeds a predetermined value and the mismatched state continues for a certain period, the value set to a main control for setting the recycling flow rate is changed corresponding to the mismatching degree to compensate the same thereby preventing the liquid level from lowering in the reactor. (Ikeda, J.)

  12. Crud removal with deep bed type condensate demineralizer in Tokai-2 BWR

    International Nuclear Information System (INIS)

    Abe, Ayumi; Takiguchi, Hideki; Numata, Kunio; Saito, Toshihiko

    1996-01-01

    The major objective and functions for the installation of the deep bed type condensate polishers in BWR power plants is to remove both ionic impurities caused by sea water leakage and suspended impurities called crud mainly consisting of metal oxides which are produced from metal corrosion. In considering the reduction of occupational radiation exposure level, it is extremely important to remove the crud effectively. In recent Japanese BWR power plants, condensate pre-filters with powdered ion exchange resins or with hollow fiber membrane have been installed to remove the crud at the upper stream of the deep bed polishers. In such plants, the crud removal is conventionally the secondary objective for the deep bed polishers. The Japan Atomic Power Company has introduced the small particle ion exchange resin and a soak regeneration method since April 1985, and then applied the low cross-linked resin since July 1995 at Tokai-2 Power Station, to improve the crud removal performance by using only deep bed type condensate demineralizer, and as a result condensate demineralizer outlet iron level has been kept below 1 ppb since 1991

  13. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  14. Status report on the application of process noise technique in nuclear power plants

    International Nuclear Information System (INIS)

    Espefaelt, R.; Aakerhielm, F.

    1979-09-01

    The report gives a survey of applications of noise technique reported for nuclear power plants. The scope has been limited to areas of interest for BWR and PWR plants of the types found in Sweden and with an emphasis on cases where the practical applicability has been clearly demonstrated. (author)

  15. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

  16. Experience gained in the training of nuclear power plant operating personnel with nuclear power plant simulators

    International Nuclear Information System (INIS)

    Buettner, J.; Fueg, J.; Schlegel, G.

    1980-01-01

    The simulator of a PWR-type reactor with 1.200 MW was accomplished in September 1977. In January 1978, the simulator of a BWR-type reactor with 800 MW started operation. The American company Singer/Link supplied computer hardware and software; Kraftwerk Union AG supplied control room equipment, power plant data and acted as consulting engineers for the construction and acceptance of the simulators. This way it is ensured that the simulated process reflects the state of German nuclear engineering. (orig./DG) [de

  17. Plant computer system in nuclear power station

    International Nuclear Information System (INIS)

    Kato, Shinji; Fukuchi, Hiroshi

    1991-01-01

    In nuclear power stations, centrally concentrated monitoring system has been adopted, and in central control rooms, large quantity of information and operational equipments concentrate, therefore, those become the important place of communication between plants and operators. Further recently, due to the increase of the unit capacity, the strengthening of safety, the problems of man-machine interface and so on, it has become important to concentrate information, to automate machinery and equipment and to simplify them for improving the operational environment, reliability and so on. On the relation of nuclear power stations and computer system, to which attention has been paid recently as the man-machine interface, the example in Tsuruga Power Station, Japan Atomic Power Co. is shown. No.2 plant in the Tsuruga Power Station is a PWR plant with 1160 MWe output, which is a home built standardized plant, accordingly the computer system adopted here is explained. The fundamental concept of the central control board, the process computer system, the design policy, basic system configuration, reliability and maintenance, CRT display, and the computer system for No.1 BWR 357 MW plant are reported. (K.I.)

  18. Complete BWR--EM LOCA analysis using the WRAP--EM system

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

    1979-01-01

    The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

  19. Safety of nuclear power plants in the 21st century

    International Nuclear Information System (INIS)

    Kovacs, Z.; Novakova, H.; Rydzi, S.

    2012-01-01

    Discussing the disaster of March 2011 which had a destroying effect on the Fukushima Dai-ichi nuclear power plant, the article presents an overview of the impacts of the earthquake and tsunami on the nuclear power plants in the region, outlines the defence-in-depth concept, and describes the design of the affected BWR type reactors and the accident event sequences leading to the reactor core damage and radioactivity release into the environment. The proposed measures for enhancing nuclear reactor safety in the 21st century are highlighted. (orig.)

  20. BWR Water Chemistry Guidelines: 1993 Revision, Normal and hydrogen water chemistry

    International Nuclear Information System (INIS)

    Karlberg, G.; Goddard, C.; Fitzpatrick, S.

    1994-02-01

    The goal of water chemistry control is to extend the operating life of the reactor and rector coolant system, balance-of-plant components, and turbines while simultaneously controlling costs to safeguard the continued economic viability of the nuclear power generation investment. To further this goal an industry committee of chemistry personnel prepared guidelines to identify the benefits, risks, and costs associated with water chemistry in BWRs and to provide a template for an optimized water chemistry program. This document replaces the BWR Normal Water Chemistry Guidelines - 1986 Revision and the BWR Hydrogen Water Chemistry Guidelines -- 1987 Revision. It expands on the previous guidelines documents by covering the economic implications of BWR water chemistry control

  1. Development of power change maneuvering method for BWR

    International Nuclear Information System (INIS)

    Fukuzaki, Takaharu; Yamada, Naoyuki; Kiguchi, Takashi; Sakurai, Mikio.

    1985-01-01

    A power change maneuvering method for BWR has been proposed to generate an optimal power control maneuver, which realizes the power change operation closest to a power change demand pattern under operating constraints. The method searches for the maneuver as an optimization problem, where the variables are thermal power levels sampled from the demand pattern, the performance index is defined to express the power mismatch between demand and feasible patterns, and the constraints are limit lines on the thermal power-core flow rate map and limits on keeping fuel integrity. The usable feasible direction method is utilized as the optimization algorithm, with newly developed techniques for initial value generation and step length determination, which apply one-dimensional search and inverse-interpolation methods, respectively, to realize the effective search of the optimal solution. Simulation results show that a typical computing time is about 5 min by a general purpose computer and the method has been verified to be practical even for on-line use. (author)

  2. Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, S.; LaChance, J.; Mary Drouin

    1995-01-01

    This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat removal are important contributors for the BWR 3/4 plants, while transients, station blackout sequences, and loss-of-coolant accidents are important for the Westinghouse 4-loop plants. The key factors that contribute to the variability in the results are discussed. The results are often driven by plant-specific design and operational characteristics, but differences in modeling approaches are also important for some accident classes

  3. Data list of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Nakamura, Jinichi

    1982-10-01

    This report has collected and compiled the data by December in 1981 concerning performances, equipments and installations of the nuclear power plants in Japan. The data have been modified according to the changes produced after previous publication of 1979 edition including BWR and PWR (JAERI-M 8947) and 1980 edition including PWR (JAERI-M 9629), and extended to cover the new plants developed thereafter. All data have been processed and tabulated with a data processing computer program FREP. Besides this report, user also can refer to 'Data List of Nuclear Power Plant in Japan' through terminals equipped at various places in JAERI using TSS (Time Shearing System) network of FACOM M-200, and the explanation of the usage is given in the Appendix. (author)

  4. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  5. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  6. BWR Servicing and Refueling Improvement Program: Phase I summary report

    International Nuclear Information System (INIS)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

  7. Intermediate size LWR plant study for process heat plus power. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    Head, M.A.

    1977-01-01

    The appropriateness of intermediate sized LWRs is evaluated for application to the process industry and for cogeneration of electric power and process steam. This brief study is directed toward determination of whether such plants show enough promise to warrant more detailed investigation. In light of higher fossil fuel costs, the study shows that intermediate sized, standardized power plants potentially are economically competitive for such industrial applications. A representative intermediate sized operating plant of the BWR/4 design class, the Swiss Muhleberg unit (1000 MWt) has been examined with respect to design, licensability, capacity factor and cost. It has operated at high capacity factor (approximately 75 percent) since turnover 11/72. Its cost when escalated from 1969 to 1976 ($620/kWe) appears competitive. Cost adjustments ($100-$250/kWe) included at this stage for compliance with current licensing and mandatory design requirements are only a preliminary estimate. Further study is recommended to confirm necessary regulatory upgrades for this BWR/4 nuclear plant and to explore specific cost economies through replication leading to a program for construction of a demonstration plant

  8. New technology for BWR power plant control and instrumentation

    International Nuclear Information System (INIS)

    Takano, Yoshiyuki; Nakamura, Makoto; Murata, Fumio.

    1992-01-01

    Nuclear power plants are facing strong demands for higher reliability and cost-performance in their control and instrumentation systems. To meet these needs, Hitachi is developing advanced control and instrumentation technology by rationalizing the conventional technology in that field. The rationalization is done through the utilization of reliable digital technology and optical transmission technology, and others, which are now commonly used in computer applications. The goal of the development work is to ensure safe, stable operation of the plant facilities and to secure harmony between man and machine. To alleviate the burdens of the operators, the latest electronic devices are being employed to create an advanced man-machine interface, and to promote automatic operation of the plant based upon the automatic operation of individual systems. In addition, the control and instrumentation system, including the safety system, incorporates more and more digital components in order to further enhance the reliability and maintainability of the plant. (author)

  9. Strategy of plant concept development in Hitachi

    International Nuclear Information System (INIS)

    Kumiaki, Moriya

    2007-01-01

    Hitachi contributes to the society in various fields of nuclear power such as the Light Water Reactor field, the Fast Breeder Reactor, the fuel cycle, and the medical treatment. Since the beginning of a first commercial operation of a BWR in Japan, Hitachi has constructed 20 units of BWR. Hitachi continues its efforts in achieving high reliability and large-scale output, and in 1996, it completed Advanced Boiling Water Reactor (ABWR) in cooperation with various BWR utilities, General Electric Company, and Toshiba Corporation. Hitachi has enhanced the ABWR technology further based on the above enough experience. The latest technologies were reflected in Hamaoka unit 5 and Shika unit 2 as the latest ABWR plants. The further upgrade technologies would been reflected in Shimane-3, Ohma-1 and Higashidori-1 as ABWR plants under planning. Hitachi obtains the chance to construct the nuclear power plant continuously. For the next generation, Hitachi is working on developing nuclear power plants that take diversified needs and global characteristics into account. As one of the approaches, the output series formation of BWR is extended as the following. ABWR-II and ESBWR (Economic and Simplified BWR) as a large-scale centralized power supply emphasizing cost efficiency, Medium size ABWR and natural circulation type BWR as medium and small-scale distributed power supply that features flexibility to various market needs, such as minimized capital risks, timely return on a capital investment, etc. As another approach, Hitachi tries to extend the Light Water Reactor technology. RBWR (Resource-Renewable BWR ) that achieves a high conversion ratio over 1.0 based on the BWR technology will make the fuel cycle flexible. Hitachi will continue the challenge for the next ABWR and the future with the enough experience of BWR construction. (author)

  10. Application of Continuous and Structural ARMA modeling for noise analysis of a BWR coupled core and plant instability event

    International Nuclear Information System (INIS)

    Demeshko, M.; Dokhane, A.; Washio, T.; Ferroukhi, H.; Kawahara, Y.; Aguirre, C.

    2015-01-01

    Highlights: • We demonstrate the first application of a novel CSARMA method. • We analyze the instability occurred in a Swiss BWR plant during power ascension. • Benchmarked the results against STP analysis. • The CSARMA results are consistent with the background physics and the STP results. • The instability was caused by disturbances in the pressure control system. - Abstract: This paper presents a first application of a novel Continuous and Structural Autoregressive Moving Average (CSARMA) modeling approach to BWR noise analysis. The CSARMA approach derives a unique representation of the system dynamics by more robust and reliable canonical models as basis for signal analysis in general and for reactor diagnostics in particular. In this paper, a stability event that occurred in a Swiss BWR plant during power ascension phase is analyzed as well as the time periods that preceded and followed the event. Focusing only on qualitative trends at this stage, the obtained results clearly indicate a different dynamical state during the unstable event compared to the two other stable periods. Also, they could be interpreted as pointing out a disturbance in the pressure control system as primary cause for the event. To benchmark these findings, the frequency-domain based signal transmission-path (STP) method is also applied. And with the STP method, we obtained similar relationships as mentioned above. This consistency between both methods can be considered as being a confirmation that the event was caused by a pressure control system disturbance and not induced by the core. Also, it is worth noting that the STP analysis failed to catch the relations among the processes during the stable periods, that were clearly indicated by the CSARMA method, since the last uses more precise models as basis

  11. Operation and maintenance support expert systems for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Miyazawa, Tatsuo; Hayakawa, Hiroyasu; Fukutomi, Shigeki

    1990-01-01

    Toshiba has been developing expert systems for operation and maintenance support in BWRs. These expert systems are designed to be integrated with conventional plant monitoring systems, and maintenance management systems to provide both comprehensive and powerful support capabilities. Some of these expert systems are already being used on a trial basis both within Toshiba and in electric power utility companies. This paper describes expert systems for the support of plant and equipment monitoring, maintenance scheduling, and maintenance work procedure planning. (author)

  12. An application of the process computer and CRT display system in BWR nuclear power station

    International Nuclear Information System (INIS)

    Goto, Seiichiro; Aoki, Retsu; Kawahara, Haruo; Sato, Takahisa

    1975-01-01

    A color CRT display system was combined with a process computer in some BWR nuclear power plants in Japan. Although the present control system uses the CRT display system only as an output device of the process computer, it has various advantages over conventional control panel as an efficient plant-operator interface. Various graphic displays are classified into four categories. The first is operational guide which includes the display of control rod worth minimizer and that of rod block monitor. The second is the display of the results of core performance calculation which include axial and radial distributions of power output, exit quality, channel flow rate, CHFR (critical heat flux ratio), FLPD (fraction of linear power density), etc. The third is the display of process variables and corresponding computational values. The readings of LPRM, control rod position and the process data concerning turbines and feed water system are included in this category. The fourth category includes the differential axial power distribution between base power distribution (obtained from TIP) and the reading of each LPRM detector, and the display of various input parameters being used by the process computer. Many photographs are presented to show examples of those applications. (Aoki, K.)

  13. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  14. Evaluation of thermal margin during BWR neutron flux oscillation

    International Nuclear Information System (INIS)

    Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

    1992-01-01

    Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

  15. Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

    International Nuclear Information System (INIS)

    Pruitt, D.W.

    1992-01-01

    This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

  16. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  17. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

  18. Reactor safety study applied to the Forsmark 3 Power Plant

    International Nuclear Information System (INIS)

    Ericsson, G.; Tiren, L.I.

    1978-01-01

    A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

  19. Surveillance and fault diagnosis for power plants in the Netherlands: operational experience

    International Nuclear Information System (INIS)

    Turkcan, E.; Ciftcioglu, O.; Hagen, T.H.J.J. van der

    1998-01-01

    Nuclear Power Plant (NPP) surveillance and fault diagnosis systems in Dutch Borssele (PWR) and Dodewaard (BWR) power plants are summarized. Deterministic and stochastic models and artificial intelligence (AI) methodologies effectively process the information from the sensors. The processing is carried out by means of methods and algorithms that are collectively referred to Power Reactor Noise Fault Diagnosis. Two main schemes used are failure detection and instrument fault detection. In addition to conventional and advanced modern fault diagnosis methodologies involved, also the applications of emerging technologies in Dutch reactors are given and examples from operational experience are presented. (author)

  20. Optimization and improvement of the technical specifications for Santa Maria de Garona and Cofrentes nuclear power plants

    International Nuclear Information System (INIS)

    Norte Gomez, M.D.; Alcantud, F.; Hoyo, C. del

    1993-01-01

    Technical Specifications (TS) form one of the basic documents necessary for licensing nuclear power plants and are required by the Government in accordance with Article 26 of the Regulation for Nuclear and Radioactive Facilities. They contain specific plant characteristics and operating limits to provide adequate protection for the safety and health of operators and the general public. For operator actuation, TS include all the surveillance requirements and limiting operating conditions (operation at full power, startup, hot and cold shutdown, and refueling outage) of safety-related systems. They also include the conventional support systems which are necessary to keep the plant in a safe operating conditioner to bring it to safe shutdown in the event of incidents or hypothetical accidents. Because of the large volume of information contained in the TS, the NRC and American utility owners began to simplify and improve the initial standard TS, which has given way to the development of a TS Optimization Program in the USA under the auspices of the NRC. Empresarios Agrupados has been contracted by the BWR Spanish Owners' Group (GPE-BWR) to develop optimized TS for the Santa Maria de Garona and Cofrentes Nuclear Power Plants. The optimized and improved TS are simplified versions of the current ones and facilitate the work of plant operators. They help to prevent risks, and reduce the number of potential transients caused by the large number of tests required by current TS. Plant operational safety is enhanced and higher effective operation is achieved. The GPE-BWR has submitted the first part of the optimized TS with their corresponding Bases to the Spanish Nuclear Council (CSN), for comment and subsequent approval. Once the TS are approved by the Spanish Nuclear Council, the operators of the Santa Maria de Garona and Cofrentes Nuclear Power Plants will be given a training and adaptation course prior to their implementation. (author)

  1. European BWR R and D cluster for innovative passive safety systems

    International Nuclear Information System (INIS)

    Hicken, E.F.; Lensa, W. von

    1996-01-01

    The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de

  2. Aging Management Guideline for commercial nuclear power plants: Battery chargers, inverters and uninterruptible power supplies

    International Nuclear Information System (INIS)

    Berg, R.; Stroinski, M.; Giachetti, R.

    1994-02-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant battery chargers, inverters and uninterruptible power supplies important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already, experienced) and aging management program activities to the more generic results and recommendations presented herein

  3. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  4. BWR containments license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Smith, S.; Gregor, F.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

  5. Nuclear power plants

    International Nuclear Information System (INIS)

    Ushijima, Susumu.

    1984-01-01

    Purpose: To enable to prevent the degradation in the quality of condensated water in a case where sea water leakage should occur in a steam condenser of a BWR type nuclear power plant. Constitution: Increase in the ion concentration in condensated water is detected by an ion concentration detector and the leaking factor of sea water is calculated in a leaking factor calculator. If the sea water leaking factor exceeds a predetermined value, a leak generation signal is sent from a judging device to a reactor power control device to reduce the reactor power. At ehe same tiem, the leak generation signal is also sent to a steam condenser selection and isolation device to interrupt the sea water pump of a specified steam condenser based on the signal from the ion concentration detector, as well as close the inlet and outlet valves while open vent and drain valves to thereby forcively discharge the sea water in the cooling water pipes. This can keep the condensate desalting device from ion breaking and prevent the degradation in the quality of the reactor water. (Horiuchi, T.)

  6. Water quality control method and device for nuclear power plant and nuclear power plant

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Uetake, Naoto; Sawa, Toshio; Uchida, Shunsuke; Takeda, Renzo; Osumi, Katsumi.

    1993-01-01

    In a BWR type nuclear power plant, water quality of coolants is controlled so as to lower deposition rate of Co ions in reactor water on a fuel cladding tube. The water quality control method includes (1) decreasing an iron concentration in feedwater to less than 0.1ppb, (2) adjusting coolants weakly acidic and (3) controlling dissolved oxygen concentration in reactor water to 20ppb. This can decrease 60 Co ion concentration even if 60 Co ion concentration is increased by the change of environment for the operation in future, such as an operation with hydrogen injection and extention of fuel burnup degree. (T.M.)

  7. Application of the neutron noise analysis technique in nuclear power plants

    International Nuclear Information System (INIS)

    Lescano, Victor H.; Wentzeis, Luis M.

    1999-01-01

    Using the neutron noise analysis in nuclear power plants, and without producing any perturbation in the normal operation of the plant, information of the vibration state of the reactor internals and the behavior of the operating conditions of the reactor primary circuit can be obtained. In Argentina, the neutron noise analysis technique is applied in customary way in the nuclear power plants Atucha I and Embalse. A database was constructed and vibration frequencies corresponding to different reactor internals were characterized. Reactor internals with particular mechanical vibrations have been detected and localized. In the framing of a cooperation project between Argentina and Germany, we participated in the measurements, analysis and modelisation, using the neutron noise technique, in the Obrigheim and Gundremmingen nuclear power plants. In the nuclear power plant Obrigheim (PWR, 350 M We), correlations between the signals measured from self-power neutron detectors and accelerometers located inside the reactor core, were made. In the nuclear power plant Gundremmingen (BWR, 1200 M We) we participated in the study of a particular mechanical vibration detected in one of the instrumentation tube. (author)

  8. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  9. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  10. 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

  11. Performances of nuclear power plants for combined production of electricity and hot water for district heating

    International Nuclear Information System (INIS)

    Bronzen, S.

    The possibilities for using nuclear power plants for combined production of heat and power seem to be very good in the future. With the chosen 600 MWsub (e) BWR plant a heat output up to 1200 MW can be arranged. An alternative, consisting of steam extractions from the low-pressure turbine, offers a flexible solution for heat and power generation. With this alternative the combined plant can use components from normal condensing nuclear power plants. The flexible extraction design also offers a real possibility for using the combined plant in electric peak generation. However, urban siting requires long distance heat transmission and the pipe design for this transmission is a major problem when planning and optimizing the whole nuclear combined heat and power plant. (author)

  12. Power plant and utility performance: how world-record outages are being achieved in the USA

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    Two record-breaking refuelling outages at power reactors in the USA are described. The first, at Browns Ferry 3 BWR, was accomplished in 19 days 39 minutes - a shorter time than for an General Electric BWR anywhere in the world hitherto. The management attribute this success to planning, personnel and performance. As well as refuelling, inspections and maintenance, major modifications were carried out. These included the completion of the installation of digital feedwater reactor level control and digital feedwater heater level control. The second outage, at South Texas Project 2 BWR, at 17 days 14 hours and 10 minutes was the fastest yet recorded for any US nuclear unit. This achievement is ascribed to excellent outage preparation and scheduling, the superior condition of the plant equipment and teamwork and safety consciousness on behalf of the plant personnel. Finally, brief consideration is given to the nuclear performance recovery programme of Commonwealth Edison and Ontario Hydro Nuclear. (UK)

  13. Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

    2009-01-01

    In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2007, TENPES published a revised edition of the guideline. This is the report of the revised edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent accumulation of radiolysis gas. (author)

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

    International Nuclear Information System (INIS)

    1995-04-01

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

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

    International Nuclear Information System (INIS)

    1995-04-01

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

  16. After-heat removing device in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mizuno, K [Nippon Atomic Industry Group Co. Ltd., Tokyo

    1977-01-14

    Purpose: To prevent water hammer in a BWR type reactor or the like by moving water in pipe lines having stagnant portions in an after-heat removing device. Constitution: To a reactor container, is provided a recycling pump which constitutes a closed loop type recycling system in a nuclear power plant together with a pressure vessel and pipe lines. A pump and a heat exchanger are provided outside of the reactor container and they are connected to up- and down-streams of the recycling pump to form an after-heat removing device in the plant. Upon shutdown of the nuclear power plant, since water in the stagnant portion flows to the intake port of the recycling pump and water from the reactor is spontaneously supplemented thereafter to the stagnant portion, neither pressurized water nor heated steam is generated and thus water hammer is prevented.

  17. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    Eckardt, B.

    1991-01-01

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO 2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  18. High-speed BWR power plant simulations on the special-purpose peripheral processor AD10

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

    1985-01-01

    A newly developed technique is described for fast, on-line simulations of normal and accidental transients in nuclear power plants. The technique is based on the utilization of the special-purpose peripheral processor AD10, which is specifically designed for high-speed systems simulations through integration of large systems of nonlinear ordinary differential equations. The Peach Bottom-II Boiling Water Reactor power plant has been simulated and results are presented. It is shown that the new technique not only advances safety analyses but also supports plant monitoring, failure diagnosis and accident mitigation, as well as the training of nuclear power plant operators. (author)

  19. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde; Estudio de ruido ambiental en una planta BWR como la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  20. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  1. Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

    International Nuclear Information System (INIS)

    Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

    2002-01-01

    Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

  2. Stress corrosion cracking of L-grade stainless steels in boiling water reactor (BWR) plants

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Fukuda, Toshihiko; Yamashita, Hironobu

    2004-01-01

    L-grade stainless steels as 316NG, SUS316L and SUS304L have been used for the BWR reactor internals and re-circulation pipes as SCC resistant materials. However, SCC of the L-grade material components were reported recently in many Japanese BWR plants. The detail investigation of the components showed the fabrication process such as welding, machining and surface finishing strongly affected SCC occurrence. In this paper, research results of SCC of L-grade stainless steels, metallurgical investigation of core shrouds and re-circulation pipings, and features of SCC morphology were introduced. Besides, the structural integrity of components with SCC, countermeasures for SCC and future R and D planning were introduced. (author)

  3. Water chemistry control practices and data of the European BWR fleet

    International Nuclear Information System (INIS)

    Stellwag, B.; Laendner, A.; Weiss, S.; Huettner, F.

    2010-01-01

    Nineteen BWR plants are in operation in Europe, nine built by ASEA Atom, six by Siemens KWU and four by General Electric. This paper gives an overview of water chemistry operation practices and parameters of the European BWR plants. General design characteristics of the plants are described. Chemistry control strategies and underlying water chemistry guidelines are summarized. Chemistry data are presented and discussed with regard to plant design characteristics. The paper is based on a contract of the European BWR Forum with AREVA on a chemistry sourcebook for member plants. The survey of chemistry data was conducted for the years 2002 to 2008. (author)

  4. Availability of thermal power plants 1985-1994. 24. ed.

    International Nuclear Information System (INIS)

    Nitsch, D.; Schmitz, H.

    1995-01-01

    The survey in hand is the 24th statistical report in the series commenced in 1970. It covers the ten-year period from 1985 through 1994 and presents availability data of 349 power plants in Germany and abroad, representing approx. 99.000 MW and a total of 3.500 years of operating experience. Data are presented on fossil-fuel units, units with a combined gas/steam cycle, nuclear power plants, and gas turbines. The fossil-fuel units are broken down by unit size, years of operation, fuel, type of combustion (dry, melt), and design type (monoblock and duoblock, subcritical and supercritical systems). Nuclear power plants are arranged by type of reactor (PWR, BWR), unit size, and years of operation. Combined-cycle power plants are listed separately due to their various technical design concepts. The gas turbine data are arranged by years of operation. Apart from availability and utilisation data of gas turbines, there are data on event reliability and the number of successful or unsuccessful starts. In general, data for all plants and systems included are given first whenever appropriate, the data for the German plants following in second place. Performance data are gross values measured at generator terminals and, just as the number of plants, are end-of-the-year figures. (orig./GL) [de

  5. Application of passive auto catalytic recombiner (PAR) for BWR plants in Japan

    International Nuclear Information System (INIS)

    Kobayashi, K.; Murano, K.; Yamanari, S.; Yamamoto, Y.

    2001-01-01

    The passive auto-catalytic recombiner (PAR), which can recombine flammable gases such as hydrogen and oxygen with each other to avoid an explosion in case of a loss-of-coolant accident (LOCA), installed in the primary containment vessel does not require a power supply or dynamic equipment, while the existing flammability gas control system (FCS) of most BWRs as an outer loop of the primary containment vessel needs them to make flammable gases circulate through blowers and heaters in the system. PAR offers a number of advantages over existing FCS, such as high reliability, low cost due to much smaller amount of materials needed, good maintainability, good operability in case of a LOCA, and smaller space for installation. An experimental study has been carried out for the purpose of solving the problems of applying PAR to Japanese BWR plants instead of existing FCS, in which we grasped the basic characteristics of PAR. (author)

  6. Total-system expertise in economically efficient operation of nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    Siemens Nuclear Power GmbH can look back on well over 40 years of experience in developing and constructing nuclear power plants. 23 Power plant units of Siemens design are in operation in five countries, and in autumn this year, another one will start commercial operation, while yet another one is under construction. In comparative international power plant surveys, the Siemens-design systems usually rank in top positions when it comes to comparing systems availability and electric power generation, and Siemens have build a reputation in manufacturing power plants up to the highest safety standards worldwide. Our experience as a manufacturer of turnkey PWR and BWR type reactors, as well as our profound knowledge of international nuclear standardisation, engineering codes and safety guides, has been used and processed to the benefit of the services offered by Siemens, resulting in well-devised service packages, and enhancements and optimisation of our machinery and equipment. Siemens has of course obtained the relevant licenses and certification for all its services and products according to DIN ISO 9001, KTA and ASME standards [de

  7. Operating the plant, quality assurance, and the job of the operating staff, Volume Twelve

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Subject matter includes operating the plant (the role of the operator, the control room, plant technical specifications, plant operating procedures, initial startup program, BWR/PWR plant startup, BWR/PWR steady state power operation, BWR/PWR transient operation, emergency operation), quality assurance (what is quality, what is quality control, quality assurance includes quality control, government regulation and quality assurance, administrative controls for nuclear power plants, the necessity of reviews and audits, practical quality assurance), and the job of the operating staff (the plant operating staff, plant safety, first aid and resuscitation, general plant hazards, personnel protective equipment, handling chemicals, handling compressed gas, equipment repair and maintenance, communicating with others

  8. LWR [Light Water Reactor] power plant simulations using the AD10 and AD100 systems

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab

  9. Nuclear power plant simulation on the AD10

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

    1985-01-01

    A combination of advanced modeling techniques and the modern, special-purpose peripheral minicomputer AD10 is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feedwater train. Point kinetics incorporate reactivity feedback for void fraction, for fuel temperature, for coolant temperature, and for boron concentration. Control systems and trip logic are simulated for the nuclear steam supply system. 4 refs., 3 figs

  10. Evaluation of robotic inspection systems at nuclear power plants

    International Nuclear Information System (INIS)

    White, J.R.; Eversole, R.E.; Farnstrom, K.A.; Harvey, H.W.; Martin, H.L.

    1984-03-01

    This report presents and demonstrates a cost-effective approach for robotics application (CARA) to surveillance and inspection work in existing nuclear power plants. The CARA was developed by the Remote Technology Corporation to systematically determine the specific surveillance/inspection tasks, worker hazards, and access or equipment placement restraints in each of the many individual rooms or areas at a power plant. Guidelines for designing inspection robotics are included and are based upon the modular arrangement of commercially-available sensors and other components. Techniques for maximizing the cost effectiveness of robotics are emphasized in the report including: selection of low-cost robotic components, minimal installation work in plant areas, portable systems for common use in different areas, and standardized robotic modules. Factors considered as benefits are reduced radiation exposure, lower man-hours, shorter power outage, less waste material, and improved worker safety concerns. A partial demonstration of the CARA methodology to the Sequoyah (PWR) and Browns Ferry (BWR) Plants is provided in the report along with specific examples of robotic installations in high potential areas

  11. Aging management guideline for commercial nuclear power plants-pumps

    International Nuclear Information System (INIS)

    Booker, S.; Katz, D.; Daavettila, N.; Lehnert, D.

    1994-03-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  12. Aging management guideline for commercial nuclear power plants-pumps

    Energy Technology Data Exchange (ETDEWEB)

    Booker, S.; Katz, D.; Daavettila, N.; Lehnert, D. [MDC-Ogden Environmental and Energy Services, Southfield, MI (United States)

    1994-03-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  13. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Porsmyr, Jan; Bodal, Terje; Beere, William H.

    2004-01-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

  14. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

    2004-07-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

  15. Nuclear power for developing countries

    International Nuclear Information System (INIS)

    Hirschmann, H.; Vennemann, J.

    1980-01-01

    The paper describes the energy policy quandary of developing countries and explains why nuclear power plants of a suitable size - the KKW 200 MW BWR nuclear power plant for electric power and/or process steam generation is briefly presented here - have an economic advantage over fossil-fuelled power plants. (HP) [de

  16. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng [Atomic Energy Council, Taoyuan City, Taiwan (China). Inst. of Nuclear Energy Research

    2017-11-15

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  17. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    International Nuclear Information System (INIS)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng

    2017-01-01

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  18. Nuclear power plant safety. The merits of separation

    International Nuclear Information System (INIS)

    Helander, L.I.; Tiren, L.I.

    1977-01-01

    The paper illustrates how the physical separation of safety-related structures, systems and components can improve the protection of a nuclear power plant against multiple failures that may be caused by events such as fire, pipe-whip, missiles, flooding, hurricanes, lightning etc. Criteria for redundancy and separation requirements affect nuclear plant design in many areas such as building layout, arrangements for fire protection and ventilation, separation of mechanical systems and components, in particular emergency cooling systems, separation of electric equipment and cables. The implementation of the ensuing design criteria for a BWR power plant is described. This design involves the separation of emergency cooling systems into four 50% capacity systems, which are independent and separated, including the distribution network for electric power from on-site standby diesel generators and the circuitry for the reactor protection system. The plant is subdivided into a number of fire zones, each with its own independent ventilation system. The fire zones are further subdivided into a multitude of fire cells such that redundant subsystems are housed in separate cells. These design precautions with regard to fire are complemented by extensive fire fighting systems. (author)

  19. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  20. Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

    1995-09-01

    The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

  1. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

  2. Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke [Energy Research Lab., Ibaraki (Japan); Ohsumi, Katsumi; Takashima, Yoshie [Hitachi Works, Ibaraki (Japan)

    1995-03-01

    Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increase must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Advanced man-machine system for nuclear power plants

    International Nuclear Information System (INIS)

    Masui, Takao; Naito, Norio; Kato, Kanji.

    1990-01-01

    Recent development of artificial intelligence(AI) seems to offer new possibility to strengthen the performance of the operator support system. From this point of view, a national project of Advanced Man-Machine System Development for Nuclear Power Plant (MMS-NPP) has been carried out since 1984 as 8-year project. This project aims at establishing advanced operator support functions which support operators in their knowledge-based behaviors and smoother interface with the system. This paper describes the role of MMS-NPP, the support functions and the main feature of the MMS-NPP detailed design with its focus placed on the realization methods using AI technology of the support functions for BWR and PWR plants. (author)

  5. Chemistry of the water in thermal power plants

    International Nuclear Information System (INIS)

    Freier, R.K.

    1984-01-01

    This textbook and practical manual gives a comprehensive review of the scientific knowledge of water as operating substance and of the chemistry of water in thermal power plants. The fundamentals of water chemistry and of the conventional and nuclear water/steam circuit are described. The contents of the chapters are: 1. The atom, 2. The chemical bond, 3. The dissolving capacity of water, 4. Operational parameters and their measurement, 5. Corrosion, 6. The water/steam coolant loop of conventional plants (WSC), 7. The pressurized water reactor (PWR), 8. The boiling water reactor (BWR), 9. The total and partial desalination properties of ion exchangers, 10. The cooling water, 11. The failure of Harrisburg in a simple presentation. (HK) [de

  6. BWR full integral simulation test (FIST) pretest predictions with TRACBO2

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.

    1984-01-01

    The Full Integral Simulation Test program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An analytical method development program is underway to extend the BWR-TRAC computer code to model reactor kinetics and major interfacing systems, including balance-of-plant, to improve application modeling flexibility, and to reduce computer running time. An experimental program is underway in a new single bundle system test facility to extend the large break loss-of-coolant accident LOCA data base to small breaks and operational transients. And a method qualification program is underway to test TRACBO2 against experiments in the FIST facility. The recently completed Phase 1 period included a series of LOCA and power transient tests, and successful pretest analysis of the large and small break LOCA tests with TRACBO2. These comparisons demonstrate BWR-TRAC capability for small and large break analysis, and provide detailed understanding of the phenomena

  7. Comparative of fuel cycle cost for light water nuclear power plants

    International Nuclear Information System (INIS)

    Kocic, A.; Dimitrijevic, Z.

    1978-01-01

    Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

  8. Impact of advanced BWR core physics method on BWR core monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

    2000-07-01

    Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

  9. Generation of artificial earthquakes for dynamic analysis of nuclear power plant

    International Nuclear Information System (INIS)

    Tsushima, Y.; Hiromatsu, T.; Abe, Y.; Tamaki, T.

    1979-01-01

    A procedure for generating artificial earthquakes for the purpose of the dynamic analysis of the nuclear power plant has been studied and relevant computer codes developed. This paper describes brieafly the generation procedure employed in the computer codes and also deals with the results of two artificial earthquakes generated as an example for input motions for the aseismic design of a BWR-type reactor building. Using one of the generated artificial earthquakes and two actually recorded earthquakes, non-linear responses of the reactor building were computed and the results were compared with each other. From this comparison, it has been concluded that the computer codes are practically usable and the generated artificial earthquakes are useful and powerful as input motions for dynamic analysis of a nuclear power plant. (author)

  10. Aging management guideline for commercial nuclear power plants-stationary batteries

    International Nuclear Information System (INIS)

    Berg, R.; Shao, J.; Krencicki, G.; Giachetti, R.

    1994-03-01

    The Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant stationary batteries important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  11. Aging Management Guideline for commercial nuclear power plants: Electrical switchgear

    International Nuclear Information System (INIS)

    Toman, G.; Gazdzinski, R.; Schuler, K.

    1993-07-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of this AMG to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner which allows personnel responsible for performance analysis and maintenance, to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  12. Advanced chemistry management system to optimize BWR chemistry control

    International Nuclear Information System (INIS)

    Maeda, K.; Nagasawa, K.

    2002-01-01

    BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

  13. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-01-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  14. Reactor power control device

    International Nuclear Information System (INIS)

    Doi, Kazuyori.

    1981-01-01

    Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

  15. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  16. Seismic resistance design of nuclear power plant building structures in Japan

    International Nuclear Information System (INIS)

    Kitano, Takehito

    1997-01-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  17. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  18. An analysis of human maintenance failures of a nuclear power plant

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-01-01

    In the report, a study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components appeared to be especially prone to human failures. Many human failures were found in safety related systems. Several failures also remained latent from outages to power operation. However, the safety significance of failures was generally small. Modifications were an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more in the future. (orig.)

  19. Statistical analysis of human maintenance failures of a nuclear power plant

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-01-01

    In this paper, a statistical study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components are especially prone to human failures. Many human failures were found in safety related systems. Similarly, several failures remained latent from outages to power operation. The safety significance was generally small. Modifications are an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more, in future. (orig.)

  20. Improvement of ALARA in Taiwan Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kao, H. M.; Chang, M.

    2006-01-01

    Full text: Full text: The ALARA(as low as reasonable achievable ) principle is a basic spirit of radiation safety culture and the most important element of a successful radiation safety program. In recent years, the ALARA principle and practices have been widely applied to the three operating nuclear power plants in Taiwan. Through the effective communication and fully cooperation between the authority and utility, the performance goals of improving radiation safety program have been set, and the implementation plans have been proposed and followed. Taking Chinshan nuclear power station as an example, in 2004, the occupational individual dose was 1.13mSv/ person in average,no one exceeded 20mSv ,and the collective dose was 1.03man-Sv/unit which achieved the best record of BWR plant in Taiwan. The radiation reduction approaches and tools adopted by the plants include removing radiation sources, setting radiation shieding, optimizing and controlling the schedule of radiation practices, considering the occupancy factors, and better administrative management to keep exposure ALARA will be discussed in this paper

  1. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

  2. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

    Borondo, L.; Serrano, C.; Fiol, M.J.; Sanchez, A.

    2000-01-01

    Cofrentes Nuclear Power Plant (GE BWR/6) has implemented its specific Severe Accident Management Program within this year 2000. New organization and guides have been developed to successfully undertake the management of a severe accident. In particular, the Technical Support Center will count on a new ''Severe Accident Management Team'' (SAMT) which will be in charge of the Severe Accident Guides (SAG) when Control Room Crew reaches the Emergency Operation Procedures (EOP) step that requires containment flooding. Specific tools and training have also been developed to help the SAMT to mitigate the accident. (author)

  3. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Johnsen, G.W.; Lellouche, G.S.

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

  4. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1982-01-01

    A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  5. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1983-01-01

    A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  6. Dry well cooling systems in BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Hanamura, Ikuo; Tada, Kenji.

    1986-01-01

    Purpose: To prevent the damages of pipeways due to salt damages at the surface of control rod drives in BWR type reactors. Constitution: In control rod drives and the lowermost area in the dry well in which surface corrosion and pitching have been resulted by the salt contents in air due to the increase in the humidity accompanying the lowering of the temperature, a blower is disposed to the upstream of the cooling coils and a portion of high temperature air returned to the lower cooler is replaced with a low temperature feed air to increase the feed temperature in the area. Further, by upwardly turning the downwarded feed air drawing port in which cold feed air has so far been descended as it is, the descendance of the cold air is suppressed. As a result, temperature lowering in the driving mechanisms and the lower area can be prevented to obtain a predetermined temperature, whereby the dewing on the surface can be prevented and thereby preventing the occurrence of corrosion and pitching. (Horiuchi, T.)

  7. Chemistry in power plants 2011

    International Nuclear Information System (INIS)

    2011-01-01

    Within the VGB Powertech conference from 25th to 27th October, 2011, in Munich (Federal Republic of Germany), the following lectures and poster contributions were presented: (1) The revised VGB standard for water-steam-cycle Chemistry; (2) Switchover from neutral operation to oxygen treatment at the power station Stuttgart-Muenster of EnBW Kraftwerke AG; (3) Steam contamination with degradation products of organic matters present in the feedwater of the Lanxess-Rubber cogeneration plant; (4) Laboratory scale on-line noble metal deposition experiments simulating BWR plant conditions; (5) Building a new demin installation for the power plant EPZ in Borssele; (6) Replacement of the cooling tower installations in the nuclear power plant Goesgen-Daenien AG; (7) Aging of IEX resins in demin plants - Cost optimisation by adaptation of regenerants; (8) The largest DOW trademark EDI System at a combined cycled plant in Europe; (9) Upgrading river Main water to boiler feed water - Experiences with ultrafiltration; (10) Experiences with treatment of the water-steam-cycle in the RDF power plant Nehlsen Stavenhagen with film-forming amines; (11) Comparative modelling of the bubbles thermal collapse and cavitations for estimation of bubbles collapse influence; (12) Overcoming the steam quality - issues from an HRSG for the production of process steam; (13) Legionella - new requirements for power plant operation; (14) How the right chemistry in the FGD helps to improve the removal in the waste water treatment plant; (15) High efficiency filtration in dry/semi-dry FGD plants; (16) Expanding the variety of renewable fuels in the biomass power plant Timelkam using the chemical input control; (17) Corrosion, operating experiences and process improvements to increase the availability and operating time of the biomass power plant Timelkam; (18) The influence of temperature on the measurement of the conductivity of highly diluted solutions; (19) A multiparameter instrumentation approach

  8. Assessment of the Prony's method for BWR stability analysis

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

    2011-01-01

    Highlights: → This paper describes a method to determine the degree of stability of a BWR. → Performance comparison between Prony's and common AR techniques is presented. → Benchmark data and actual BWR transient data are used for comparison. → DR and f results are presented and discussed. → The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

  9. Cavitation preventing device in a nuclear power plant

    International Nuclear Information System (INIS)

    Hirose, Masao.

    1983-01-01

    Purpose: To prevent the generation of cavitation upon loss of feedwater flow rate in BWR nuclear power plant by reliably and rapidly tripping a recycling pump. Constitution: Two phase streams from a nuclear reactor are divided into main steams and saturated water in a steam drum. The deviation between the corresponding flow rate of the main steams and the feedwater flow rate of the feedwater pump sending condensates to the steam drum, as well as the continuing period of the deviation are monitored. Then, if it is detected that both of the deviation and the continuing period thereof exceed specified levels, the recycling pump feeding the saturated water to the reactor is tripped. In this way, the recycling pump can be tripped rapidly and reliably upon loss of feedwater flow rate, whereby the generation of the cavitation can be prevented and the normal operation of the nuclear power plant can be insured. (Moriyama, K.)

  10. Development of a dynamic model of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Nonboel, E.

    1975-12-01

    A description is given of a one-dimensional steady-state model of a high-pressure steam turbine, a low-pressure steam turbine, a moisture separator, a reheater, a condenser, feedwater heaters and feedwater pump for a nuclear power plant. The model is contained in the program ''TURBPLANT''. The dynamic part of this model is presented in part II of this report. (author)

  11. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  12. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    International Nuclear Information System (INIS)

    Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu

    1999-01-01

    This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

  13. An engineer-constructor's view of nuclear power plant safety

    International Nuclear Information System (INIS)

    Landis, J.W.; Jacobs, S.B.

    1984-01-01

    At SWEC we have been involved in the development of safety features of nuclear power plants ever since we served as the engineer-constructur for the first commerical nuclear power station at Shippingport, Pennsylvania, in the 1950s. Our personnel have pioneered a number of safety innovations and improvements. Among these innovations is the subatmospheric containment for pressurized water reactor (PWR) power plants. This type of containment is designed so that leakage will terminate within 1 to 2 hours of the worst postulated loss of coolant accident. Other notable contributions include first use of reinforced-concrete atmospheric containments for PWR power plants and of reinforced-concrete, vapor-suppression containments for boiling water reactor (BWR) power plants. Both concepts meet rigorous U.S. safety requirements. SWEC has performed a substantial amount of work on developing standardized plant designs and has developed standardized engineering and construction techniques and procedures. Standardization concepts are being developed in Canada, France, USSR, and Germany, as well as in the United States. The West German convoy concept, which involves developing a number of standardized plants in a common effort, has been quite successful. We believe standardization contributes to safety in a number of ways. Use of standardized designs, procedures, techniques, equipment, and methods increases efficiency and results in higher quality. Standardization also reduces the design variations with which plant operators, emergency teams, and regulatory personnel must be familiar, thus increasing operator capability, and permits specialized talents to be focused on important safety considerations. (orig./RW)

  14. Systematic methodology for diagnosis of water hammer in LWR power plants

    International Nuclear Information System (INIS)

    Safwat, H.H.; Arastu, A.H.; Husaini, S.M.

    1990-01-01

    The paper gives the dimensions of the knowledge base that is necessary to carry out a diagnosis of water hammer susceptibility/root cause analyses for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear power plant systems. After introducing some fundamentals, water hammer phenomena are described. Situations where each phenomenon is encountered are given and analytical models capable of simulating the phenomena are referenced. Water hammer events in operating plants and their inclusion in the knowledge base is discussed. The diagnostic methodology is presented through an application on a system in a typical light water reactor plant. The methodology presented serves as a possible foundation for the creation of an expert water hammer diagnosis system. (orig.)

  15. ETGAR - Information system for abnormal occurrences in nuclear power plants

    International Nuclear Information System (INIS)

    Baram, J.; Nagar, M.; Pultorak, G.

    1975-01-01

    The need for extensive information on systems and components arises early in the planning stage of a nuclear power plant. This information is equally necessary during the building of the plant and during the licensing process. Another type of information helps preventive maintenance during the operating life of the plant. In the case of abnormal occurrences additional information on their possible consequences and on possible ways of handling them, is essential. To cover these four needs, the ETGAR system, which at present covers mostly PWR and BWR type nuclear power plants, collects and evaluates information on abnormal occurrences in nuclear power plants. The information is coded, using a three-level coding scheme for systems and components, and put on magnetic tape. A search program enables the retrieval of any pertinent information from the data base. The sources for the ETGAR data base are reports on abnormal occurrences in nuclear power plants. Most of them are USAEC dockets, originated at U.S.A. power plants. The relevant documents are accessible through a standard query run for ETGAR in the INIS data base which is maintained by the INIS centre in Israel. This query retrieves every two weeks all the documents which come under the ETGAR scope and these are handed as microfiches to the ETGAR evaluators after each INIS run. The evaluation and coding of the documents, the ETGAR coding scheme and the computer programs are described. (B.G.)

  16. Correlating activity incorporation with properties of oxide films formed on material samples exposed to BWR and PWR coolants in Finnish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P. [VTT Industrial Systems, Espoo (Finland); Buddas, T.; Halin, M.; Kvarnstroem, R.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant, Loviisa (Finland); Helin, M.; Muttilainen, E.; Reinvall, A. [Teollisuuden Voima Oy, Olkiluoto (Finland)

    2002-07-01

    The extent of activity incorporation on primary circuit surfaces in nuclear power plants is connected to the chemical composition of the coolant, to the corrosion behaviour of the material surfaces and to the structure and properties of oxide films formed on circuit surfaces due to corrosion. Possible changes in operational conditions may induce changes in the structure of the oxide films and thus in the rate of activity incorporation. To predict these changes, experimental correlations between water chemistry, oxide films and activity incorporation, as well as mechanistic understanding of the related phenomena need to be established. In order to do this, flow-through cells with material samples and facilities for high-temperature water chemistry monitoring have been installed at Olkiluoto unit 1 (BWR) and Loviisa unit 1 (PWR) in spring 2000. The cells are being used for two major purposes: To observe the changes in the structure and activity levels of oxide films formed on material samples exposed to the primary coolant. Correlating these observations with the abundant chemical and radiochemical data on coolant composition, dose rates etc. collected routinely by the plant, as well as with high-temperature water chemistry monitoring data such as the corrosion potentials of relevant material samples, the redox potential and the high-temperature conductivity of the primary coolant. We describe in this paper the scope of the work, give examples of the observations made and summarize the results on oxide films that have been obtained during one full fuel cycle at both plants. (authors)

  17. Method of estimating thermal power distribution of core of BWR type reactor

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1982-01-01

    Purpose: To accurately and rapidly predict the thermal power of the core of a BWR they reactor at load follow-up operating time. Method: A parameter value corrected from a correction coefficient deciding unit and a xenon density distribution value predicted and calculated from a xenon density distributor are inputted to a thermal power distribution predicting devise, the status amount such as coolant flow rate or the like predetermined at this and next high power operating times is substituted for physical model to predict and calculate the thermal power distribution. The status amount of a nuclear reactor at the time of operating in previous high power corresponding to the next high power operation to be predicted is read from the status amount of the reactor stored in time series manner is a reactor core status memory, and the physical model used in the prediction and calculation of the thermal power distribution at the time of next high power operation is corrected. (Sikiya, K.)

  18. Two types of a passive safety containment for a near future BWR with active and passive safety systems

    International Nuclear Information System (INIS)

    Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

    2009-01-01

    The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

  19. Two types of a passive safety containment for a near future BWR with active and passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Takashi [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)], E-mail: takashi44.sato@glb.toshiba.co.jp; Akinaga, Makoto; Kojima, Yoshihiro [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)

    2009-09-15

    The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

  20. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  1. 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

  2. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  3. BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

    International Nuclear Information System (INIS)

    Pohl, P.; Wehle, F.

    2010-01-01

    Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

  4. Radioactive contamination of Danish territory after core-melt accidents at the Barsebaeck power plant

    International Nuclear Information System (INIS)

    Gjoerup, H.L.; Jensen, N.O.; Hedemann Jensen, P.; Kristensen, L.; Nielsen, O.J.; Petersen, E.L.; Petersen, T.; Roed, J.; Thykier-Nielsen, S.; Heikel Vinter, F.; Warming, L.; Aarkrog, A.

    1982-03-01

    An assessment is made of the radioactive contamination of Danish territory in the event of a core-melt accident at the Barsebaeck nuclear power plant in Sweden. Accidents including both core melt-down and containment failure are considered. Consequences are calculated for a BWR-3 release under common meteorological conditions and for a BWR-2 release under extreme meteorological conditions. Calculations are based on experiments and theoretical work relating to deposition velocities for different types of surface, shielding effect of structures, and weathering. The effects are described of different dose-reducing measures, e.g., decontamination, relocation, destruction of contaminated foodstuffs. The collective effective dose equivalent from external gamma radiation from deposited activity integrated over a time period of 30 years, is calculated to be 3.6 Megamanrem in the BWR-3 case without dose-reducing measures. For the BWR-2 case, the corresponding dose is approx. 41 Megamanrem. A combination of temporary relocation, hosing of roads etc. and digging of gardens is estimated to reduce these doses to approx. 2.5 Megamanrem and approx. 15 Megamanrem, respectively. The collective committed effective dose equivalent from the consumption of contaminated foodstuffs is calculated to 23 Megamanrem in the BWR-3 case without dose-reducing measures. This dose could be reduced to 0.2 Megamanrem if contaminated crops are destroyed during the first year after the accident and if changes are made in agricultural production in the contaminated area. The corresponding doses in the BWR-2 case would be 197 Megamanrem and 1.4 Megmanrem, respectively. (author)

  5. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

  6. Seismic re-evaluation of the Tarapur atomic power plants 1 and 2

    International Nuclear Information System (INIS)

    Ingole, S.M.; Kumar, B.S.; Gupta, S.; Singh, U.P.; Giridhar, K.; Bhawsar, S.D.; Samota, A.; Chhatre, A.G.; Dixit, K.B.; Bhardwaj, S.A.

    2004-01-01

    Two Boiling Water Reactors (BWR) of 210 MWe each at Tarapur Atomic Power Station, Units-1 and 2 (TAPS-1-2) were commissioned in the year 1969. The safety related civil structures at TAPS had been designed for a seismic coefficient of 0.2 g and other structures for 0.1 g. The work of seismic re-evaluation of the TAPS-1-2 has been taken up in the year 2002. As two new Pressurized Heavy Water Reactor (PHWR) plants of 540 MWe each, Tarapur Atomic Power Project Units-3 and 4 (TAPP-3-4), are coming up in the vicinity of TAPS-1-2, detailed geological and seismological studies of the area around TAPS-1-2 are available. The same free-field ground motion as generated for TAPP-3-4 has been used for TAPS-1-2. The seismic re-evaluation of the plant has been performed as per the procedure given in IAEA, Safety Reports Series entitled 'Seismic Evaluation of Existing Nuclear Power Plants', and meeting the various codes and standards, viz., ASME, ASCE, IEEE standards etc. The Safety Systems (SS) and Safety Support Systems (SSS) are qualified by adopting detailed analysis and testing methods. The equipment in the SS and SSS have been qualified by conducting a walk-down as per the procedure given in Generic Implementation Procedure, Dept. of Energy (GIP--DOE), USA. The safety systems include the systems required for safe shutdown of the plant, one chain of decay heat removal and containment of activity. The safety support systems viz., Electrical, Instrumentation and Control and systems other than SS and SSS have been qualified by limited analysis, testing and mostly by following the procedure of walk-down. The paper brings out the details of the work accomplished during seismic re-evaluation of the two units of BWR at Tarapur. (authors)

  7. Experience for plant monitoring design in Italian BWR NPP and future trends in man-machine interface

    International Nuclear Information System (INIS)

    Maestri, F.; Sepielli, M.

    1987-01-01

    TMI accidental sequence and daily-gained operating experience on italian and abroad NPPs have affected in depth the approach to the design of information presentation to the Control Room staff. It has been cleared that most problems in plant operation arise from a poor and inadequate information system. The main lacks have been identified in the Control Room lay-out and information organization. This has pushed designers both to improve the Control Room environment and to better exploit the computer data processing and data presentation capabilities. The paper deals with the basic criteria for the design and the design review of a computerized system to be inserted in a hybrid Control Room in Italian 981 Mwe BWR-6 NPP, where the concepts outlined above were taken-up from the very beginning. The Control Room keeps conventional instrumentation arranged in a human-factor lay-out, according to post-TMI requirements, and adds a powerful computer-based information system for advanced alarm presentation and plant supervision during both normal and emergency conditions with high data reliability. Colour videounits and operating panels are functionally integrated to create powerful operator work-stations. Emphasis is mostly given on the revision work for video-unit displays and Man-System Communication carried out in cooperation with Halden Reactor Project human factor and plant operation experts. The work peculiarity has been a strong care on the integration between conventional and computerized information presentation, with particular regard to common information and code consistency. (author)

  8. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  9. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  10. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions

  11. Analysis of natural circulation BWR dynamics with stochastic and deterministic methods

    International Nuclear Information System (INIS)

    VanderHagen, T.H.; Van Dam, H.; Hoogenboom, J.E.; Kleiss, E.B.J.; Nissen, W.H.M.; Oosterkamp, W.J.

    1986-01-01

    Reactor kinetic, thermal hydraulic and total plant stability of a natural convection cooled BWR was studied using noise analysis and by evaluation of process responses to control rod steps and to steamflow control valve steps. An estimate of the fuel thermal time constant and an impression of the recirculation flow response to power variations was obtained. A sophisticated noise analysis method resulted in more insight into the fluctuations of the coolant velocity

  12. Indicators for management of planned outages in nuclear power plants

    International Nuclear Information System (INIS)

    2006-04-01

    The outages considered within the scope of this publication are planned refuelling outages (PWR and BWR nuclear power plants) and planned outages associated with major maintenance, tests and inspections (PHWR and LWGR nuclear power plants). The IAEA has published some valuable reports providing guidance and assistance to operating organizations on outage management. This TECDOC outlines main issues to be considered in outage performance monitoring and provides guidance to operating organizations for the development and implementation of outage programmes which could enhance plant safety, reliability and economics. It also complements the series of reports published by the IAEA on outage management and on previous work related to performance indicators developed for monitoring different areas of plant operation, such as safety, production, reliability and economics. This publication is based upon the information presented at a technical meeting to develop a standardized set of outage indicators for outage optimization, which was organised in Vienna, 6-9 October 2003. At this meeting, case studies and good practices relating to performance indicator utilization in the process of planned outage management were presented and discussed

  13. Concrete containment vessels (CCV) for nuclear power plants, (1)

    International Nuclear Information System (INIS)

    Ibe, Yukimi; Kitajima, Masatake

    1977-01-01

    Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

  14. Evaluating and improving nuclear power plant operating performance

    International Nuclear Information System (INIS)

    1999-07-01

    This report aims to provide the basis for improvements in the understanding of nuclear power plants operation and ideas for improving future productivity. The purpose of the project was to identify good practices of operating performance at a few of the world's most productive plants. This report was prepared through a series of consultants meetings, a specialists meeting and an Advisory Group meeting with participation of experts from 23 Member States. The report is based on self-assessment of half a dozen plants that have been chosen as representatives of different reactor types in as many different countries, and the views and assessment of the participants on good practices influencing plant performance. Three main areas that influence nuclear power plant availability and reliability were identified in the discussions: (1) management practices, (2) personnel characteristics, and (3) working practices. These areas cover causes influencing plant performance under plant management control. In each area the report describes factors or good practices that positively influence plant availability. The case studies, presented in annexes, contain the plant self-assessment of areas that influence their availability and reliability. Six plants are represented in the case studies: (1) Dukovany (WWER, 1760 MW) in the Czech Republic; (2) Blayais (PWR, 3640 MW) in France; (3) Paks (WWER, 1840 MW) in Hungary; (4) Wolsong 1 (PHWR, 600 MW) in the Republic of Korea; (5) Trillo 1 (PWR, 1066 MW) in Spain; and (6) Limerick (BWR, 2220 MW) in the United States of America

  15. Fuel performance experience at TVO nuclear power plant

    International Nuclear Information System (INIS)

    Patrakka, E.T.

    1985-01-01

    TVO nuclear power plant consists of two BWR units of ASEA-ATOM design. The fuel performance experience extending through six cycles at TVO I and four cycles at TVO II is reported. The experience obtained so far is mainly based on ASEA-ATOM 8 x 8 fuel and has been satisfactory. Until autumn 1984 one leaking fuel assembly had been identified at TVO I and none at TVO II. Most of the problems encountered have been related to leaf spring screws and channel screws. The experience indicates that satisfactory fuel performance can be achieved when utilizing strict operational rules and proper control of fuel design and manufacture. (author)

  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. Rock siting of nuclear power plants from a reactor safety standpoint

    International Nuclear Information System (INIS)

    1975-11-01

    The study has aimed at surveying the advantages and disadvantages of a rock sited nuclear power plant from a reactor safety standpoint. The studies performed are almost entirely concentrated on the BWR alternative. The design of a nuclear power plant in rock judged most appropriate has been studied in greater detail, and a relatively extensive safety analysis has been made. It is found that the presented technical design of the rock sited alternative is sufficiently advanced to form a basis for further projecting treatment. The chosen technical design of the reactor plant demands a cavern with a 45-50 metre span. Caverns without strengthening efforts with such spans are used in mines, but have no previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant. When locating a rock sited nuclear power plant, the same criteria must be considered as for an above ground plant, with additional stronger demands for rock quality. The presented rock sited nuclear power plant has been assessed to cost 20 % more in total construction costs than a corresponding above ground plant. The motivations for rock siting also depend on whether a condensing plant for only electricity production, or a plant for combined power production and district heating, is considered. The latter would under certain circumstances make rock siting look more attractive. (author)

  18. Aging management and preventive maintenance for nuclear power plants

    International Nuclear Information System (INIS)

    Bessho, Toichi; Sagawa, Wataru; Oyamada, Osamu; Uchida, Shunsuke

    1995-01-01

    It is expected that nuclear power generation will bear main electric power supply for long term. For this purpose, by applying proper preventive maintenance to the nuclear power plants in operation, the maintenance of high reliability and the rate of plant operation is extremely important. Especially it has been strongly demanded to execute efficiently the periodic inspection which is carried out every year to shorten its period and increase the rate of operation, and to maintain the reliability by the proper maintenance for the aged plants with long operation years. As to efficient and short periodic inspection, the preparation is advanced by the guidance of electric power companies aiming at realizing it in nearest fiscal year, and further, effort is exerted for the development of technology in long term to optimize periodic inspection. For securing the reliability of aged plants, it is important to do proper inspection and diagnosis and to take the countermeasures by repair and replacement, based on the grasp of secular change mechanism and the evaluation of life of machinery, equipment and materials. In particular, effort is exerted for the development of maintenance technology for reactor pressure vessels and in-core equipment which are hard to access. The confirmation of the function of remote operation equipment and the establishment of execution condition are carried out by uisng the full scale mock-up of BWR plant. The problems in maintenance service and the activities and results of reliability improvement are reported. (K.I.)

  19. FEMAXI-7 analysis on behavior of medium and high burnup BWR fuels during base-irradiation and power ramp

    Energy Technology Data Exchange (ETDEWEB)

    Ogiyanagi, Jin, E-mail: ohgiyanagi.jin@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. Black-Right-Pointing-Pointer Calculated FGR and cladding deformation showed reasonable agreement with PIE data. Black-Right-Pointing-Pointer High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. Black-Right-Pointing-Pointer Local PCMI model in the code could reasonably predict cladding ridging deformation. - Abstract: Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.

  20. Facility for processing the condensates from nuclear power plants (BWR and PWR)

    International Nuclear Information System (INIS)

    Lucker, Georges.

    1975-01-01

    A plant for the processing of the condensates from boiling water or pressurized water nuclear power plants is presented. A series of couples of units for the processing of the condensates through mixed beds of ion exchange resins simultaneously ensures the filtration and demineralization of the condensates. When the resins are saturated, each mixed bed is transferred into a unit of regeneration of said resins. Each processing unit is a sphere made of a stainless material, and provided with a plurality of air and water pipes allowing the admission and evacuation of the various elements to be successively controlled [fr

  1. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  2. Aging Management Guideline for commercial nuclear power plants: Motor control centers

    International Nuclear Information System (INIS)

    Toman, G.; Gazdzinski, R.; O'Hearn, E.

    1994-02-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  3. LWR-plants. Their evolutionary progress in the last half-century. (4) The start of the nuclear power generation in Japan

    International Nuclear Information System (INIS)

    Ishikawa, Hiroshi

    2008-01-01

    Evolutionary progress of the LWR plants in the last half-century was reviewed in series. The start of the nuclear power generation in Japan was reviewed in this article. The Japan Atomic Energy Research Institute (JAERI) promoted nuclear power research and development and introduced the Japan Power Demonstration Reactor (JPDR)-a 12.5 MWe natural circulation BWR, which began operation in 1965. In 1969 its power uprate modifications to a forced circulation BWR (JPDR-II) began and attained operation in 1972. During 50% power test, primary coolant leakage was observed at reactor core spray pipes in 1972. In 1975 the operation resumed and faults observed at condenser tubes in 1976. Primary coolant leakage from in-core flux monitor guide tubes at the bottom of reactor pressure vessel in 1979 led to its permanent shutdown. The nuclear ship Mutsu was put into service in 1970 and during rising power test radiation leakage due to fast neutron streaming was observed in 1974. After modifications of shielding experimental voyage was made in 1991. The first commercial nuclear power reactor, Tokai-1-a 166 MWe gas-cooled (Magnox) reactor, began operation in 1966 and continued until 1998. The LWR plants became the mainstay in Japan. (T. Tanaka)

  4. System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

    2009-10-15

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  5. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    International Nuclear Information System (INIS)

    Adamsson, Carl; Le Corre, Jean-Marie

    2011-01-01

    Highlights: → The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. → A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. → MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. → The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. → The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the

  6. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  7. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    International Nuclear Information System (INIS)

    Kmetyk, L.N.; Brown, T.D.

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP ampersand S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP ampersand S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP ampersand S configuration are given

  8. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

  9. Development of a BWR loading pattern design system based on modified genetic algorithms and knowledge

    International Nuclear Information System (INIS)

    Martin-del-Campo, Cecilia; Francois, Juan Luis; Avendano, Linda; Gonzalez, Mario

    2004-01-01

    An optimization system based on Genetic Algorithms (GAs), in combination with expert knowledge coded in heuristics rules, was developed for the design of optimized boiling water reactor (BWR) fuel loading patterns. The system was coded in a computer program named Loading Pattern Optimization System based on Genetic Algorithms, in which the optimization code uses GAs to select candidate solutions, and the core simulator code CM-PRESTO to evaluate them. A multi-objective function was built to maximize the cycle energy length while satisfying power and reactivity constraints used as BWR design parameters. Heuristic rules were applied to satisfy standard fuel management recommendations as the Control Cell Core and Low Leakage loading strategies, and octant symmetry. To test the system performance, an optimized cycle was designed and compared against an actual operating cycle of Laguna Verde Nuclear Power Plant, Unit I

  10. Application of expert system to nuclear power plant operation and guidance system

    International Nuclear Information System (INIS)

    Goto, M.; Takada, Y.

    1990-01-01

    For a nuclear power plant, it is important that an expert system supplies useful information to the operator to meet the increasing demand for high-level plant operation. It is difficult to build a user-friendly expert system that supplies useful information in real time using existing general-purpose expert system shells. Therefore a domain-specific expert system shell with a useful knowledge representation for problem-solving in nuclear power plant operation was selected. The Plant Table (P/T) representation format was developed for description of a production system for nuclear power plant operation knowledge. The P/T consists of plant condition representation designed to process multiple inputs and single output. A large number of operation inputs for several plant conditions are divided into 'timing conditions', 'preconditions' and 'completion conditions' to facilitate knowledge-base build-up. An expert system for a Nuclear Power Plant Operation and Guidance System utilizing the P/T was developed to assist automatic plant operation and surveillance test operation. In these systems, automatic plant operation signals to the plant equipment and operation guidance messages to the operators are both output based on the processing and assessment of plant operation conditions by the P/T. A surveillance test procedure guide for major safety-related systems, such as those for emergency core cooling systems, is displayed on a CRT (Cathode Ray Tube) and test results are printed out. The expert system for a Nuclear Power Plant Operation and Guidance System has already been successfully applied to Japanese BWR plants

  11. An estimation of population doses from a nuclear power plant during normal operation

    International Nuclear Information System (INIS)

    Nowicki, K.

    1975-07-01

    A model is presented for estimation of the potential submersion and inhalation radiation doses to people located within a distance of 1000 km from a nuclear power plant during normal operation. The model was used to calculate doses for people living 200-1000 km from hypothetical nuclear power facility sited near the geographical centre of Denmark. Two kinds of sources are considered for this situation: - unit release of 15 isotopes of noble gases and iodines, - effluent releases from two types of 1000 MWe Light Water Power Reactors: PWR and BWR. Parameter variations were made and analyzed in order to obtain a better understanding of the mechanisms of the model. (author)

  12. TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kozlowski, T. [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, T.; Xu, Y.; Wysocki, A. [Univ. of Michigan, Ann Arbor, MI (United States); Ivanov, K.; Magedanz, J.; Hardgrove, M. [Pennsylvania State Univ., Univ. Park, PA (United States); March-Leuba, J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hudson, N.; Woodyatt, D. [Nuclear Regulatory Commission, Rockville, MD (United States)

    2012-07-01

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)

  13. Start up and commercial operation of Laguna Verde nuclear power plant. Unit 1

    International Nuclear Information System (INIS)

    Torres Ramirez, J.F.

    1991-01-01

    Prior to start up of Laguna Verde nuclear power plant preoperational tests and start tests were performed and they are described in its more eminent aspects. In relation to commercial operation of nuclear station a series of indicator were set to which allow the measurement of performance in unit 1, in areas of plant efficiency and personal safety. Antecedents. Laguna Verde station is located in Alto Lucero municipality in Veracruz state, 70 kilometers north-northeast from port of Veracruz and a 290 kilometers east-northeast from Mexico city. The station consist of two units manufactured by General Electric, with a nuclear system of vapor supply also called boiling water (BWR/5), and with a system turbine-generator manufactured by Mitsubishi. Each unit has a nominal power of 1931 MWt and a level design power of 675 Mwe and a net power of 654 Electric Megawatts

  14. Direct cycle type nuclear power plant

    International Nuclear Information System (INIS)

    Tagawa, Hisato; Ibe, Hidefumi.

    1990-01-01

    In a direct cycle type nuclear power plant such as BWR type reactor, since oxygen atoms in reactor water are actuvated by neutron irradiation in the reactor core, carry over of the thus formed radioactive nitrogen atoms causes increase in the dosage in a turbine system. Since 16 N accompanies in the main steams in the chemical form of 16 NO, it can not effectively be removed in a nitrogen removing device. In view of the above, hydrogen atom concentration is reduced by adding metals having high reaction with hydrogen atoms, for example, silver ions, chromium ions, or ruthenium ions are added to reactor water. Then, equilibrium concentration of 16 NO in water is reduced by suppressing the reaction: 16 NO 2 + H → 16 NO + OH. (T.M.)

  15. Present situation of floating nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, A [Central Research Inst. of Electric Power industry, Tokyo (Japan)

    1975-08-01

    The present situation of investigation and the future problems of floating nuclear power plants in Japan are examined, referring to those in USA. The committee report on a new power generation system in 1970 is quoted. In this report, the site conditions are supposed to be 5 km offshore, 100m water depth, 60 m/sec wind velocity, 10 m wave height, 200 m wave length, 12 seconds wave period 0.2 g earthquake acceleration, and 2.5 knots tide current. The semisubmersible hull of double construction 15 m under water is employed. A pair of 1,000,000 kW BWR reactors are utilized. A sea water desalting unit using bleed steam from turbines is installed. The solid radioactive wastes packed in drums are disposed in the sea. The design and cost estimation were made. The names of the organizations who have made investigation in this field, namely the Civil Engineering Society, the Sience and Technology Agency and other several centers, are reported. The Chubu Electric Power Company is forwarding its project. Referring to the investigations in USA, the project of Atlantic nuclear power station unit is described. A report of plant design has been submitted by O.P.S. to United States Atomic Energy Commission in 1973. The Coastal Area Facilities Act was instituted in New Jersey in 1973. Although the Atlantic nuclear power station has been postponed, it is the most feasible project. For the realization of a floating nuclear power plant in Japan, investigation must be started on the ground construction that can endure the construction of breakwater in water depth of 14 to 30 meter.

  16. Control rod pattern exchange in a BWR/6 utilizing gang mode withdrawal

    International Nuclear Information System (INIS)

    Auvil, A.B. Jr.; Aldemir, T.; Hajek, B.K.

    1986-01-01

    The use of checkerboard pattern of alternating inserted and fully withdrawn control rods and the uneven void distribution in boiling water reactor (BWR) cores can cause large burnup gradients even after a short time of operation. To compensate for these effects, power has to be reshaped periodically (typically every two full-power months) by individually manipulating the control rods. During this manipulation process (called the control rod pattern exchange), the core power is reduced to 60% of nominal power by means of flow reduction to limit power swings to tolerable levels and to ensure that fuel thermal limits are not exceeded. A control rod pattern exchange by individual rod manipulation typically takes 4 to 8 h and represents a large cost burden to the utility in terms of reduced system output. The latest generation of BWRs, the BWR/6, possesses the capability to simultaneously move up to four symmetrically located control rods. The rods corresponding to a given gang may have rotational symmetry, mirror symmetry, or a combination of the two. This paper presents a pattern exchange procedure that exploits the capability of gang mode rod withdrawal to reduce the pattern exchange execution time and radial power distribution asymmetry associated with individual rod manipulation. The working model used in the study is the Perry Nuclear Power Plant Unit 1, located in Perry, Ohio, and owned by the Cleveland Electric Illuminating Company

  17. Pathways and estimated consequences of radionuclide releases from a nuclear power plant

    International Nuclear Information System (INIS)

    Nilsson, Mats.

    1981-03-01

    The radioactive corrosion products 60 Co, 58 Co, 54 Mn, 65 Zn, 51 Cr and 110 Agsp(m) which are released from Barsebaeck Nuclear Power Plant (Sweden) into the marine environment have been studied. The brown seaweed Fucus vesiculosus has been found to be excellent bioindicator for these radionuclides. The distribution of activation products along the west coast of Sweden has been studied and can very well be described with a power function. Uptake and retention of the activation products in Fucus have been studied and results are reported. The activity concentration in Fucus well reflects the disharge rate from the power plant. Other bioindicators have been studied ; mainly the crustaceans Idothea and Gammarus. A simulation of 131 I-release following a fictious BWR 1 accident is described as well as a model for estimation of the individual and collective dose equivalents arising from intake of contaminated milk. (Author)

  18. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael

    2012-09-01

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  19. Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plant:2002

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Suzuki, Riichiro; Komura, Seiichi; Kudo, Yoshiro; Yamanaka, Akihiro; Oomizu, Satoru; Kitamura, Hideya; Nagata, Yoshifumi

    2003-01-01

    To secure fuel integrity, a Light Water Reactor (LWR) core is designed so that no boiling transition (BT) should take place in fuel assemblies and excessive rise in fuel cladding temperature due to deteriorated that transfer should be avoided in Anticipated Operational Occurrences (AOO). In some AOO in a Boiling Water Reactor (BWR), however, the rise in reactor power could be limited by SCRAM or void reactivity effect. Recent studies have provided accumulated knowledge that even if BT takes place in fuel assemblies, the rise in fuel cladding temperature could be so small that it will not threat to fuel integrity, as long as the BT condition terminates within a short period of time. In addition, appropriate methods have been developed to evaluate the cladding temperature during dryout. This standard provides requirements in the assessment of fuel integrity under AOO in which limited-BT condition is temporarily reached and the propriety of reusing a fuel assembly that has experienced limited-BT condition. The standard has been approved by the Atomic Energy Society of Japan following deliberation by impartial members for two and half years. It is now expected that this standard will provide an effective measure for the rational expansion of fuel design and operational margin. (author)

  20. Evaluating and improving nuclear power plant operating performance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    This report aims to provide the basis for improvements in the understanding of nuclear power plants operation and ideas for improving future productivity. The purpose of the project was to identify good practices of operating performance at a few of the world`s most productive plants. This report was prepared through a series of consultants meetings, a specialists meeting and an Advisory Group meeting with participation of experts from 23 Member States. The report is based on self-assessment of half a dozen plants that have been chosen as representatives of different reactor types in as many different countries, and the views and assessment of the participants on good practices influencing plant performance. Three main areas that influence nuclear power plant availability and reliability were identified in the discussions: (1) management practices, (2) personnel characteristics, and (3) working practices. These areas cover causes influencing plant performance under plant management control. In each area the report describes factors or good practices that positively influence plant availability. The case studies, presented in annexes, contain the plant self-assessment of areas that influence their availability and reliability. Six plants are represented in the case studies: (1) Dukovany (WWER, 1760 MW) in the Czech Republic; (2) Blayais (PWR, 3640 MW) in France; (3) Paks (WWER, 1840 MW) in Hungary; (4) Wolsong 1 (PHWR, 600 MW) in the Republic of Korea; (5) Trillo 1 (PWR, 1066 MW) in Spain; and (6) Limerick (BWR, 2220 MW) in the United States of America Figs, tabs

  1. Water quality maintaining device of power plant

    International Nuclear Information System (INIS)

    Kobayashi, Minoru; Inami, Ichiro.

    1994-01-01

    The device of the present invention reduces the amount of leaching materials of ion exchange resins from a water processing system of a BWR tyep plant, improves the water quality of reactor water to maintain the water at high purity. That is, steams used for power generation are condensated in a condensate system. A condensate filter and a condensate desalter for cleaning the condensates are disposed. A resin storage hopper is disposed for supplying the ion exchange resins to the water processing system. A device for supplying a nitrogen gas or an inert gas is disposed in the hopper. With such a constitution, the ion exchange resins in the water processing system are maintained in a nitrogen gas or inert gas atmosphere or at a low dissolved oxygen level in an operation stage in the power plant. Accordingly, degradation of the ion exchange resins in the water processing system is suppressed and the amount of the leaching material from the resins is reduced. As a result, the amount of the resins leached into the reactor is reduced, so that the reactor water quality can be maintained at high purity. (I.S.)

  2. ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: ROSA-III is a 1/124 scaled down test facility with electrically heated core designed to study the response of engineered safety features to loss-of-coolant accidents in in commercial BWR. It consists of the following, fully instrumented subsystems: (a) the pressure vessel with a core simulating four half-length fuel assemblies and control rod; (b) steam line and feed water line, which are independent open loops; (c) coolant recirculation system, which consists of two loops provided with a recirculation pump and two jet pumps in each loop; (d) emergency cooling system, including HPCS, LPCS, LPCI, and ADS. 2 - Description of test: Run 971 simulated a BWR LOSS of off-site power transient. The core scram was assumed to occur at 6 seconds after the transient initiated by the turbine trip. HPCS failure was assumed. After ADS started, the upper half of the core was uncovered by steam. The core was re-flooded by LPCS alone

  3. Investigation of BWR stability in Forsmark 2

    International Nuclear Information System (INIS)

    Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

    1988-01-01

    A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  4. ECP measurements in the BWR-1 water loop relative to water composition changes

    Energy Technology Data Exchange (ETDEWEB)

    Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2010-07-01

    The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

  5. In-Containment Signal Conditioning and Transmission via Power Lines within High Dose Rate Areas of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Steffen; Weigel, Robert; Koelpin, Alexander [Institute for Electronics Engineering, University of Erlangen-Nuremberg, Cauerstr. 9, 91058 Erlangen (Germany); Dennerlein, Juergen; Janke, Iryna; Weber, Johannes [AREVA GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)

    2015-07-01

    Signal conditioning and transmission for sensor systems and networks within the containment of nuclear power plants (NPPs) still poses a challenge to engineers, particularly in the case of equipment upgrades for existing plants, temporary measurements, decommissioning of plants, but also for new builds. This paper presents an innovative method for efficient and cost-effective instrumentation within high dose rate areas inside the containment. A transmitter-receiver topology is proposed that allows simultaneous, unidirectional point-to-point transmission of multiple sensor signals by superimposing them on existing AC or DC power supply cables using power line communication (PLC) technology. Thereby the need for costly installation of additional cables and containment penetrations is eliminated. Based on commercial off-the-shelf (COTS) electronic parts, a radiation hard transmitter is designed to operate in harsh environment within the containment during full plant operation. Hardware modularity of the transmitter allows application specific tradeoffs between redundancy and channel bandwidth. At receiver side in non-radiated areas, signals are extracted from the power line, demodulated, and provided either in analog or digital output format. Laboratory qualification tests and field test results within a boiling water reactor (BWR) are validating the proof of concept of the proposed system. (authors)

  6. Development status of compact containment BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

    2005-01-01

    In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response

  7. Technology and costs for decommissioning of Swedish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-06-01

    The decommissioning study for the Swedish nuclear power plants has been carried out during 1992 to 1994 and the work has been led by a steering group consisting of people from the nuclear utilities and SKB. The study has been focused on two reference plants, Oskarshamn 3 and Ringhals 2. Oskarshamn 3 is a boiling water reactor (BWR) and Ringhals 2 is a pressurized water reactor (PWR). Subsequently, the result from these plants have been translated to the other Swedish plants. The study gives an account of the procedures, costs, waste quantities and occupational doses associated with decommissioning of the Swedish nuclear power plants. Dismantling is assumed to start immediately after removal of the spent fuel. No attempts at optimization, in terms of technology or costs, have been made. The nuclear power plant site is restored after decommissioning so that it can be released for use without restriction for other industrial activities. The study shows that a reactor can be dismantled in about five years, with an average labour force of about 150 persons. The maximum labour force required for Oskarshamn 3 has been estimated to about 300 persons. This peak load occurred the first years but is reduced to about 50 persons during the demolishing of the buildings. The cost of decommissioning Oskarshamn 3 has been estimated to be about MSEK 940 in January 1994 prices. The decommissioning of Ringhals 2 has been estimated to be MSEK 640. The costs for the other Swedish nuclear power plants lie in the range MSEK 590-960. 17 refs, 21 figs, 15 tabs.

  8. Technology and costs for decommissioning of Swedish nuclear power plants

    International Nuclear Information System (INIS)

    1994-06-01

    The decommissioning study for the Swedish nuclear power plants has been carried out during 1992 to 1994 and the work has been led by a steering group consisting of people from the nuclear utilities and SKB. The study has been focused on two reference plants, Oskarshamn 3 and Ringhals 2. Oskarshamn 3 is a boiling water reactor (BWR) and Ringhals 2 is a pressurized water reactor (PWR). Subsequently, the result from these plants have been translated to the other Swedish plants. The study gives an account of the procedures, costs, waste quantities and occupational doses associated with decommissioning of the Swedish nuclear power plants. Dismantling is assumed to start immediately after removal of the spent fuel. No attempts at optimization, in terms of technology or costs, have been made. The nuclear power plant site is restored after decommissioning so that it can be released for use without restriction for other industrial activities. The study shows that a reactor can be dismantled in about five years, with an average labour force of about 150 persons. The maximum labour force required for Oskarshamn 3 has been estimated to about 300 persons. This peak load occurred the first years but is reduced to about 50 persons during the demolishing of the buildings. The cost of decommissioning Oskarshamn 3 has been estimated to be about MSEK 940 in January 1994 prices. The decommissioning of Ringhals 2 has been estimated to be MSEK 640. The costs for the other Swedish nuclear power plants lie in the range MSEK 590-960. 17 refs, 21 figs, 15 tabs

  9. Nuclear power: 2006 world report - evaluation

    International Nuclear Information System (INIS)

    Anon.

    2007-01-01

    Last year, 2006, 437 nuclear power plants were available for power supply in 31 countries, 7 plants less than in 2005. One unit was commissioned for the first time, 8 nuclear power plants were decommissioned for good in 2006. At a cumulated gross power of 389,488 MWe and a cumulated net power of 370,441 MWe, respectively, worldwide nuclear generating capacity has reached a high level so far. Nine different reactor lines are operated in commercial plants: PWR, PWR-VVER, BWR, CANDU, D 2 O PWR, GCR, AGR, LWGR, and LMFBR. Light water reactors (PWR and BWR) continue to top the list with 358 plants. By the end of the year, 10 countries operated 29 nuclear power plants with an aggregate gross power of 25,367 MWe and an aggregate net power of 23,953 MWe, respectively. Of these, 21 are light water reactors, 5 are CANDU-type reactors, 2 are fast breeder and 1 a LWGR. 123 commercial reactors with an aggregate power in excess of 5 MWe have so far been decommissioned in 19 countries. Most of them are prototype plants of low power. About 70% of the nuclear power plants in operation, namely 304 plants, were commissioned in the eighties and nineties. The energy availability and operating availability factors of the nuclear power plants again reached peak levels: 82% for energy availability, and 83% for operating availability. The 4 nuclear power plants in Finland continue to be in the lead worldwide with a cumulated average operating capacity factor of 94%. (orig.)

  10. Nuclear power. 2008 world report - evaluation

    International Nuclear Information System (INIS)

    Anon.

    2010-01-01

    In 2008, 438 nuclear power plants were available for power supply in 31 countries, 1 plant less than in 2007. No unit was commissioned for the first time, 1 nuclear power plant was decommissioned for good in 2008. At a cumulated gross power of 392,597 MWe and a cumulated net power of 372,170 MWe, respectively, worldwide nuclear generating capacity has reached a high level. Nine different reactor lines are operated in commercial plants: PWR, PWR-VVER, BWR, CANDU, D2O PWR, GCR, AGR, LWGR, and LMFBR. Light water reactors (PWR and BWR) continue to top the list with 358 plants. By the end of 2008, in 14 countries 43 nuclear power plants with an aggregate gross power of 39,211 MWe and an aggregate net power of 36,953 MWe were under construction. Of these, 37 are light water reactors, 3 are CANDU-type reactors, 2 are fast breeder and 1 D2O-PWR. 124 commercial reactors with an aggregate power in excess of 5 MWe have so far been decommissioned in 19 countries. Most of them are prototype plants of low power. About 70% of the nuclear power plants in operation, namely 304 plants, were commissioned in the eighties and nineties. The energy availability and operating availability factors of the nuclear power plants reached good levels: 80.80% for operating availability and 80,00% for energy availability. The four nuclear power plants in Finland continuecontinue to be in the lead worldwide with a cumulated average operating capacity factor of 91,60%. (orig.)

  11. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  12. Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fifth Workshop (BWR-TT5)

    International Nuclear Information System (INIS)

    2003-01-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fifth workshop was to discuss the results from Phase III (best

  13. Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fourth Workshop (BWR-TT4)

    International Nuclear Information System (INIS)

    2002-01-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fourth workshop was to present and discuss final results of

  14. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

    Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

    1998-02-01

    The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

  15. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  16. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  17. Features of Tsuruga-2 plant

    International Nuclear Information System (INIS)

    Suzuki, Hideaki

    1984-01-01

    The Japan Atomic Power Co. was established in 1957 as the vanguard to establish commercial nuclear power plants in Japan, therefore, the approach was to build and operate one each of the commercial size nuclear power plants of latest types. In 1966, the Tokai Power Station with a gas-cooled reactor of 166 MWe imported from UK started operation. In 1970, the Tsuruga Power Station with a first BWR of 357 MWe in Japan was operated. The Tokai No. 2 Power Station with a BWR of 1100 MWe was built to introduce a large scale nuclear power plant to Japan. Now, No. 2 plant of PWR type is constructed in the Tsuruga Power Station. This plant is produced totally in Japan as the improved and standardized plant planned by the Ministry of International Trade and Industry. As the history shows, in the early stage, the repair and improvement were the major tasks, but the improvement of operation, reliability and capacity ratio, the rationalization of regular inspection, the reduction of the radiation exposure dose of workers, and the improvement of waste treatment have been gradually achieved. The model plan with four loops and 1100 MWe capacity according to the second improvement and standardization, the new technology adopted, the reflection of operation experience, and the construction method are reported. (Kako, I.)

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

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

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

  19. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant

    International Nuclear Information System (INIS)

    Zamora R, L.; Medina F, A.

    1999-01-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  20. Assessment of RELAP5/Mod3 system thermal hydraulic code using power test data of a BWR6 reactor

    International Nuclear Information System (INIS)

    Lee, M.; Chiang, C.S.

    1997-01-01

    The power test data of Kuosheng Nuclear Power Plant were used to assess RELAP5/Mod3 system thermal hydraulic analysis code. The plant employed a General Electric designed Boiling Water Reactor (BWR6) with rated power of 2894 MWth. The purpose of the assessment is to verify the validity of the plant specific RELAP5/Mod3 input deck for transient analysis. The power tests considered in the assessment were 100% power generator load rejection, the closure of main steam isolation valves (MSIVs) at 96% power, and the trip of recirculation pumps at 68% power. The major parameters compared in the assessment were steam dome pressure, steam flow rate, core flow rate, and downcomer water level. The comparisons of the system responses predicted by the code and the power test data were reasonable which demonstrated the capabilities of the code and the validity of the input deck. However, it was also identified that the separator model of the code may cause energy imbalance problem in the transient calculation. In the assessment, the steam separators were modeled using time-dependent junctions. In the approach, a complete separation of steam and water was predicted. The system responses predicted by RELAP5/Mod3 code were also compared with those from the calculations of RETRAN code. When these results were compared with the power test data, the predictions of the RETRAN code were better than those of RELAP5/Mod3. In the simulation of 100% power generator load rejection, it was believed that the difference in the steam separator model of these two codes was one of the reason of the difference in the prediction of power test data. The predictions of RELAP/Mod3 code can also be improved by the incorporation of one-dimensional kinetic model. There was also some margin for the improvement of the input related to the feedwater control system. (author)

  1. Mechanized ultrasonic examination of piping systems in nuclear power plants

    International Nuclear Information System (INIS)

    Edelmann, X.; Pfister, O.; Allidi, F.

    1988-01-01

    The success of mechanized ultrasonic examination applied on welds in piping systems in nuclear power plants is highly dependent on its careful preparation. From the development of an adequate examination technique to its implementation on site, many problems are to be solved. This is especially the case when dealing with austenitic welds or dissimilar metal welds. In addition to the specific needs for examination technique based on material properties and requirements for minimum flaw size detection, accessibility and radiation aspects have to be considered. A crew of skilled and highly trained examination personnel is required. Experience in various nuclear power plants, - BWR's and PWR's of different designs - has shown, that even difficult examination problems can be successfully solved, provided that there is a good preparation. The necessary step by step proceeding is illustrated by examples concerning mechanized examination. Preservice inspections and in-service inspections with specific requirements, due to the types of flaws to be found or the type of material concerned, are discussed

  2. Summary of Fukushima Dai-ni nuclear power plant No. 1 and by operation

    International Nuclear Information System (INIS)

    Murayama, Mamoru; Idesawa, Masato

    1984-01-01

    Fukushima No.2 Nuclear Power Station is located on the Pacific coast of Fukushima Prefecture, and the total area is about 1.5 million m 2 including about 200,000 m 2 of reclaimed land. No. 1 plant started the commercial operation on April 20, 1982, and continued the operation for 384 days. The first regular inspection was carried out for 111 days from May 9, 1983, and the second cycle operation was begun on September 13, 1983. This plant is a BWR plant of 1,100 MWe class, manufactured for the first time in Japan by Toshiba based on the BWR-5 standard design of GE. The auxiliary machine cooling system with intermediate fresh water pools, the canned motors for coolant purifying system, the automation of service equipment, long life type in-core neutron instrumentation, the analog trip circuit for the safety protection system, filtration type condensate desalting equipment, the low cobalt material for feed heater tubes and control rod pin rollers, and titanium condenser tubes are the features of this plant. Also the countermeasures to stress corrosion cracking, the improvement of the central control board, the adoption of the control rods with followers and the fuel of improved design, the concentrated treatment of radioactive wastes and so on were carried out. The results of operation are reported. (Kako, I.)

  3. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor; Analisis del accidente de la planta nucleoelectrica de Fukushima Daiichi en un reactor tipo ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Escorcia O, D. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Salazar S, E., E-mail: daniel.escorcia.ortiz@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2016-09-15

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  4. Nuclear power plant safety, the merits of separation

    International Nuclear Information System (INIS)

    Helander, L.I.; Tiren, L.I.

    1977-01-01

    The United States AEC General Design Criteria for Nuclear Power Plants are used worldwide as a basis for the assessment of nuclear plant safety. Several of these criteria require redundancy of safety systems, separation of protection and control systems, consideration of natural phenomena, etc. All these criteria point in one particular direction: the necessity for physically separating the various safety-related systems of a nuclear power plant, particularly with regard to single occurrences that may yield a multiple failure. Requirements in this regard have been amplified by the United States NRC Regulatory Guides and by IEEE Standards. The single occurrence that yields a multiple failure may be, for example, fire, pipe whip, missiles, flooding, hurricanes, or lightning. The paper discusses protection, against the quoted events and others, obtained through applying criteria regarding redundancy and separation of safety-related structures, systems and components. Such criteria affect nuclear plant design in many areas, such as building lay-out, arrangements for fire protection and ventilation, separation of mechanical systems and components, in particular emergency cooling systems, and separation of electric equipment and cables. Implementation of the ensuing design criteria for a BWR power plant is described. This design involves the separation of Emergency Cooling Systems into four 50% Capacity Systems which are independent and separated, including the distribution network for electric power from on-site standby diesel generators and the circuitry for the reactor protection system. The plant is subdivided into a number of fire zones each with its own independent ventilation system. The fire zones are further subdivided into a multitude of fire cells such that redundant subsystems are housed in separate cells. These design precautions with regard to fire are complemented by extensive fire fighting systems

  5. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    Reparaz, A.; Smith, M.H.; Stephens, L.G.

    1992-01-01

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  6. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  7. RETRAN experience with BWR transients at Yankee Atomic Electric Company

    International Nuclear Information System (INIS)

    Ansari, A.A.F.; Cronin, J.T.; Slifer, B.C.

    1981-01-01

    Yankee Atomic Electric Company is actively involved in the development of licensing methods for BWR's. The computer code chosen for analyzing system response under transient conditions is RETRAN. This paper describes the RETRAN model developed for Vermont Yankee, and the results of the RETRAN checkout and qualification that has been achieved at YAEC through comparison of RETRAN predictions to the startup test results performed at the plant as part of the 100% power startup test program. In addition, abnormal operational transients typically analyzed for licensing are also presented

  8. Application of EASY5 and MMS modules to BWR controller design

    International Nuclear Information System (INIS)

    Carmichael, L.A.; Rayes, L.; Yasutake, T.

    1987-01-01

    The application of EPRI's MMS Library and BCS' EASY5 simulation language to the design of a digital feedwater control system for the Monticello Boiling Water Nuclear Power Plant is discussed. In order to first design and then verify the digital feedwater controller algorithms, a digital simulation model of the Monticello plant was constructed using a combination of custom designed modules, existing MMS two-phase library modules, and standard modules available in the EASY5 library. Details of the process models, namely the BWR nuclear steam supply system, the steamline piping, and the feedwater piping are described in a companion paper. Details of the models for the existing BWR turbine pressure inlet pressure control and recirculation flow control system are described. These models are required to be operational during the transient analysis portion of the feedwater controller design verification, since they interact strongly with the reactor steam flow and water level. The design of the digital feedwater flow control loop is described. Its design is of particular interest because it requires consideration of control loop interaction and is, therefore, a simple example of multivariable non-interacting control design

  9. Development of remote automatic equipment for BWR power plants

    International Nuclear Information System (INIS)

    Sasaki, Masayoshi

    1984-01-01

    The development of remote control, automatic equipment for nuclear power stations has been promoted to raise the rate of operation of plants by shortening regular inspection period, to improve the safety and reliability of inspection and maintenance works by mechanization, to reduce the radiation exposure dose of workers and to reduce the manpower required for works. The taking-off of control rod drives from reactors and fixing again have been mechanized, but the disassembling, cleaning, inspection and assembling of control rod drives are manually carried out. Therefore, Hitachi Ltd. has exerted effort to develop the automatic equipment for this purpose. The target of development, investigation, the construction and function of the equipment, the performance and the effect of adopting it are reported. The equipment for the volume reduction of spent fuel channel boxes and spent control rods is developed since these are major high level radioactive solid wastes, and their apparent volume is large. Also the target of development, investigated things, the construction and function of the equipment, the performance and the effect of adopting it are reported. (Kako, I.)

  10. Fuel rod response to BWR power oscillations during anticipated transient without scram

    International Nuclear Information System (INIS)

    Cunningham, M.; Scott, H.

    1998-01-01

    The US NRC is examining fuel behaviour during a postulated BWR anticipated transient without scram (ATWS) with power oscillations to determine if current regulatory criteria are adequate. Currently, the 280 cal/g limit for RIAs is used to show that coolable geometry is maintained and pressure pulses are avoided during ATWSs. Two specific questions have now been raised about the continued use of the 280 cal/g value. First, this value was derived from energy deposition values whereas the regulatory requirements are written in terms of fuel enthalpy. The second is that fuel rod rupture with fuel dispersal has been observed in RIA tests with high bum-up fuel rods having energy deposition values well below the current limit. However, the BWR ATWS power oscillation transient is slower than a RIA power pulse, thus reducing the likelihood of failure. Therefore questions about the adequacy of the 280 cal/g limit do not necessarily imply unacceptable fuel damage occurring during such power oscillations and there is no immediate safety concern. The reported analysis, using the FRAPTRAN transient fuel rod analysis code, was thus undertaken to determine if further investigation might be appropriate and with the intention of starting some discussions about the issue. There was a comment that a limit of 100 cal/g fuel enthalpy had been mentioned following the scoping calculations but that perhaps enthalpy was not the main concern in an ATWS. It was also observed that cladding stresses are lower than in all RIA. The question was what really is the main concern. It was replied that the main concern was a question of maintaining a coolable geometry i.e. not loosing fuel particles out of the rod. And it was agreed that enthalpy may not be the important issue, rather that it previously had been used as the parameter and so had been considered. Confirmation of this presently being an evaluation and not a regulatory concern was sought and provided, it being pointed out that the NRC

  11. Collection and characterization of emissions from metal cutting in Caorso nuclear power plant decommissioning

    International Nuclear Information System (INIS)

    Cesari, F.G.; Terzi, L.A.; Giostri, A.; Bernini, C.; Sirito, E.; Sirito, M.

    2005-01-01

    The Caorso's Nuclear Power Plant (BWR, 870 MWe) has just started the decommissioning process, with the intent to reach, by almost ten years, the 'green field' conditions for the site. The plant has fully worked for a very short period of time, by 1981 to 1986, being shut down after 1987 Italy's poll that abrogated nuclear power use. The dismantling of the components and of the structural materials has already begun in the Turbine Building. The University of Bologna, on indication of the NPP's management, has started an experimental campaign to test the cutting processes and its filtering plant. The starting phase is the qualification of the cutting methods chosen by Caorso's management, oxyfuel and plasma cuttings. This campaign is set over no contaminated material, or, better, material with a contamination under the level of free release, and is now running in the University Labs. Next phase are filtering tests. This part of the qualifying campaign is set to highlight the kind of trouble that can emerge in the cutting processes, not yet taking into account radioactivity. Caorso's BWR is a plant designed and built in the 70s. Possible decommissioning problems weren't considered during the design phase so the cutting processes will be quite difficult, even on the conventional side. The final phase is settled in the plant. Cutting tests, following indications made by previous campaigns, will be conducted in the Turbine Building, where tests can be conducted on a low level of radioactivity (only some little part of it has a contamination level over the free release limit of 1 Bq/cm 2 ). The intent is to develop an extensive cutting procedure, with the obvious option of remote control, able to face difficulties connected with cutting processes in a nuclear plant like Caorso. This means handling with radioactivity and with not airy narrow rooms, fulfilled with pipes. (authors)

  12. Technical Specification action statements requiring shutdown. A risk perspective with application to the RHR/SSW systems of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy, Espoo (Finland); Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States)

    1993-11-01

    When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements.

  13. Logical model for the control of a BWR turbine;Modelo logico para el control de una turbina de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R., E-mail: yonaeton@hotmail.co [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2009-07-01

    In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

  14. System control model of a turbine for a BWR

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A.

    2009-10-01

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  15. TRACE/PARCS validation for BWR stability based on OECD/NEA Oskarshamn-2 benchmark

    International Nuclear Information System (INIS)

    Kozlowski, T.; Roshan, S.; Lefvert, T.; Downar, T.; Xu, Y.; Wysocki, A.; Ivanov, K.; Magedanz, J.; Hardgrove, M.; Netterbrant, C.; March-Leuba, J.; Hudson, N.; Sandervag, O.; Bergman, A.

    2011-01-01

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event, which culminated in diverging power oscillations with decay ratio greater than 1.3. The event was successfully modeled by TRACE/PARCS coupled code system and the details of the modeling and solution are described in the paper. The obtained results show excellent agreement with the plant data, capturing the entire behavior of the transient including onset of instability, growth of oscillation (decay ratio) and the oscillation frequency. The event allows coupled code validation for BWR with a real, challenging stability event, which challenges accuracy of neutron kinetics (NK), thermal-hydraulics (TH) and TH/NK coupling. The success of this work has demonstrated the ability of 3-D coupled code systems to capture the complex behavior of BWR stability events. The problem is released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (author)

  16. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  17. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  18. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  19. Two-loop feed water control system in BWR plants

    International Nuclear Information System (INIS)

    Omori, Takashi; Watanabe, Takao; Hirose, Masao.

    1982-01-01

    In the process of the start-up and shutdown of BWR plants, the operation of changing over feed pumps corresponding to plant output is performed. Therefore, it is necessary to develop the automatic changeover system for feed pumps, which minimizes the variation of water level in reactors and is easy to operate. The three-element control system with the water level in reactors, the flow rate of main steam and the flow rate of feed water as the input is mainly applied, but long time is required for the changeover of feed pumps. The two-loop feed control system can control simultaneously two pumps being changed over, therefore it is suitable to the automatic changeover control system for feed pumps. Also it is excellent for the control of the recirculating valves of feed pumps. The control characteristics of the two-loop feed water control system against the external disturbance which causes the variation of water level in reactors were examined. The results of analysis by simulation are reported. The features of the two-loop feed water control system, the method of simulation and the evaluation of the two-loop feed water control system are described. Its connection with a digital feed water recirculation control system is expected. (Kako, I.)

  20. Designing for nuclear power plant maintainability and operability

    International Nuclear Information System (INIS)

    Pedersen, T.J.

    1998-01-01

    Experience has shown that maintenance and operability aspects must be addressed in the design work. ABB Atom has since long an ambition of achieving optimised, overall plant designs, and efficient feedback of growing operating experience has stepwise eliminated shortcomings, and yielded better and better plant operating performances. The records of the plants of the latest design versions are very good; four units in Sweden have operated at an energy availability of 90.1%, and the two Olkiluoto units in Finland at a load factor of 92.7%, over the last decade. The occupational radiation exposures have also been at a low level. The possibilities for implementing 'lessons learned' in existing plants are obviously limited by practical constraints. In Finland and Sweden, significant modernisations are still underway, however, involving replacement of mechanical equipment, and upgrading and backfitting of I and C systems on a large scale, in most of the plants. The BWR 90 design focuses on meeting requirements from utilities as well as new regulatory requirements, with a particular emphasis on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimisation of buildings and containment to decrease construction time and costs, and selection of materials as well as maintenance of operating procedures to reduce radiation exposures even further. The BWR 90 design was offered to Finland in the early 1990s, but development work continues. It has been selected by a number of European utilities for assessing its conformance with the European Utility Requirements (EUR), aiming at a specific EUR Volume 3 for the BWR 90. Some characteristics of the ABB BWRs, with emphasis on features of importance for achieving improved economy and enhanced safety, are described below. (author)

  1. Corrosion issues in the BWR and their mitigation for plant life extension

    International Nuclear Information System (INIS)

    Gordon, B.M.

    1988-01-01

    Corrosion is a major service life limiting mechanism for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). For the BWR, stress corrosion cracking of piping has been the major source of concern where extensive research has led to a number of qualified remedies and currently > 90% of susceptible welds have been mitigated or replaced. Stress corrosion cracking of reactor internals due to the interaction of irradiation, as discussed elsewhere in this conference, is also a possible life limiting phenomenon. This paper focusses on two corrosion phenomena in the BWR which have only recently been identified as impacting the universal goal of BWR life extension: the general corrosion of containment structures and the erosion-corrosion of carbon steel piping

  2. BWR Refill-Reflood Program. Final report

    International Nuclear Information System (INIS)

    Myers, L.L.

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

  3. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    International Nuclear Information System (INIS)

    Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

    2008-01-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  4. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

    2008-07-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  5. The use of ferritic materials in light water reactor power plants

    International Nuclear Information System (INIS)

    Marston, T.V.

    1984-01-01

    This paper reviews the use of ferritic materials in LWR power plant components. The two principal types of LWR systems, the boiling water reactor (BWR) and the pressurized water reactor (PWR) are described. The evolution of the construction materials, including plates and forgings, is presented. The fabrication process for both reactors constructed with plates and forgings are described in detail. Typical mechanical properties of the reactor vessel materials are presented. Finally, one critical issue radiation embrittlement dealing with ferritic materials is discussed. This has been one of the major issues regarding the use of ferritic material in the construction of LWR pressure vessels

  6. Economic analysis of hydride fueled BWR

    International Nuclear Information System (INIS)

    Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

    2009-01-01

    The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

  7. 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)

  8. Managing BWR plant life extension

    International Nuclear Information System (INIS)

    Ianni, P.W.; Kiss, E.

    1985-01-01

    Recent studies have confirmed that extending the useful life of a large nuclear plant can be justified with very high cost benefit ratio. In turn, experience with large power plant systems and equipment has shown that a well-integrated and -managed plan is essential in order to achieve potential economic benefits. Consequently, General Electric's efforts have been directed at establishing a life extension plan that considers alternative options and cost-effective steps that can be taken in early life, those appropriate during middle life, and those required in late life. This paper briefly describes an approach designed to provide the plant owner a maximum of flexibility in developing a life extension plan

  9. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

    International Nuclear Information System (INIS)

    Pantoja C, R.

    2010-01-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  10. Effectiveness of a large mimic panel in an existing nuclear power plant central control board

    International Nuclear Information System (INIS)

    Kubota, Ryuji; Satoh, Hiroyuki; Sasajima, Katsuhiro; Kawano, Ryutaro; Shibuya Shinya

    1999-01-01

    We conducted the analysis of the nuclear power plant (NPP) operators' behaviors under emergency conditions by using training simulators as a joint research project by Japanese BWR groups for twelve years. In the phase-IV of this project we executed two kinds of experiments to evaluate the effectiveness of the interfaces. One was for evaluations of the interfaces such as CRTs with touch screen, a large mimic panel, and a hierarchical annunciator system introduced in the newly developed ABWR type central control board. The other was that we analyzed the operators' behaviors in emergency conditions by using the first generation BWR type central control board which was added new interfaces such as a large display screen and demarcation on the board to help operators to understand the plant. The demarcation is one of the visual interface improvements and its technique is that a line enclosing several components causes them to be perceived as a group.The result showed that both the large display panel Introduced in ABWR central control board and the large display screen in the existing BWR type central control board improved the performance of the NPP operators in the experiments. It was expected that introduction of the large mimic panel into the existing BWR type central control boards would improve operators' performance. However, in the case of actual installation of the large display board into the existing central control boards, there are spatial and hardware constraints. Therefore the size of lamps, lines connecting from symbols of the pumps or valves to the others' will have to be modified under these constraints. It is important to evaluate the displayed information on the large display board before actual installation. We made experiments to solve these problems by using TEPCO's research simulator which is added a large mimic panel. (author)

  11. FIREDATA, Nuclear Power Plant Fire Event Data Base

    International Nuclear Information System (INIS)

    Wheelis, W.T.

    2001-01-01

    1 - Description of program or function: FIREDATA contains raw fire event data from 1965 through June 1985. These data were obtained from a number of reference sources including the American Nuclear Insurers, Licensee Event Reports, Nuclear Power Experience, Electric Power Research Institute Fire Loss Data and then collated into one database developed in the personal computer database management system, dBASE III. FIREDATA is menu-driven and asks interactive questions of the user that allow searching of the database for various aspects of a fire such as: location, mode of plant operation at the time of the fire, means of detection and suppression, dollar loss, etc. Other features include the capability of searching for single or multiple criteria (using Boolean 'and' or 'or' logical operations), user-defined keyword searches of fire event descriptions, summary displays of fire event data by plant name of calendar date, and options for calculating the years of operating experience for all commercial nuclear power plants from any user-specified date and the ability to display general plant information. 2 - Method of solution: The six database files used to store nuclear power plant fire event information, FIRE, DESC, SUM, OPEXPER, OPEXBWR, and EXPERPWR, are accessed by software to display information meeting user-specified criteria or to perform numerical calculations (e.g., to determine the operating experience of a nuclear plant). FIRE contains specific searchable data relating to each of 354 fire events. A keyword concept is used to search each of the 31 separate entries or fields. DESC contains written descriptions of each of the fire events. SUM holds basic plant information for all plants proposed, under construction, in operation, or decommissioned. This includes the initial criticality and commercial operation dates, the physical location of the plant, and its operating capacity. OPEXPER contains date information and data on how various plant locations are

  12. Observations of crud deposits, corrosion and erosion of BWR and PWR fuel

    International Nuclear Information System (INIS)

    Bairiot, H.

    1983-01-01

    The BWR experience is limited to one reactor but the PWR experience covers a wide range of successive generations of power plants (7 in total). The systems are described and their water chemistry briefly commented. Some R and D performed on the effects of the operating regimes (steady state and transients) are summarized. Observations made by pool-side inspections and postirradiation examinations of fuel are outlined concerning water chemistry effects (crud deposits and corrosion) and ''mechanical'' coolant-cladding interaction (chip deposits and baffle jetting). (author)

  13. Controlling chemistry parameters in nuclear reactors and power plants, plant chemistry specification requirements and compliance - an overview of TAPS 1 and 2 experiences

    International Nuclear Information System (INIS)

    Ravindranath; Muralidharan, K.; Save, C.B.; Patil, D.P.

    2006-01-01

    Tarapur Atomic Power Station -TAPS 1 and 2 is a twin unit Boiling Water Reactor (BWR) Nuclear Power Plant commissioned in the year 1969. Both units are running with capacity factor of more than 90 % in their 20 th cycle of operation as on today. The 220 MWe units were derated to 160 MWe during 1984 consequent to isolation of Secondary Steam Generators (SSG) in the 10 th cycle of operation due to SSG tube leaks. This paper presents an overview of Plant Chemistry Control measures and experiences during the last 38 years of operation. The overall plant chemistry performance of TAPS 1 and 2 observed is very good; which is evident from the material condition of various systems reflected in QC and I reports, NDT and ISI reports. This is also supported by the fact that both Units are showing excellent performance continuously during recent years. (author)

  14. On results of measurement and method of behavior analysis for land slide protection wall in excavation works for main building foundation of No.2 plant in Kashiwazaki-Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc

    International Nuclear Information System (INIS)

    Otsuki, Hideo; Tsuchi, Hiroyuki; Nishioka, Toshimichi

    1987-01-01

    Tokyo Electric Power Co. has been constructing the nuclear power station having 8 million kW capacity of seven BWR plants in the site of about 4.2 million m 2 in Niigata Prefecture. No.1 BWR plant of 1100 MWe output started the operation in September, 1985. As a rule, the important structures in nuclear power stations such as a reactor building and a turbine building are to be directly supported on bedrocks, and in this case, on the mudstone of Nishiyama strata. As this Nishiyama strata exists in large depth, the excavation works for the foundations of buildings are to be carried out by installing large scale land slide protection walls. In this report, among the excavation works for the main building foundation of No.2 plant, the results of examining the behavior of the land slide protection wall installed in soft rock ground based on the results of measurement of vertical excavation by land slide protection method and the techniques of its analysis are described. The geological features, the design of land slide protection walls, the measurement of the land slide protection walls and surrounding ground and the results, and the examination of the analysis methods by a beam model and FEM are reported. (Kako, I.)

  15. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

    2014-01-01

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

  16. A BWR 24-month cycle analysis using multicycle techniques

    International Nuclear Information System (INIS)

    Hartley, K.D.

    1993-01-01

    Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

  17. Improvement of nuclear power plant monitor and control equipment. Computer application backfitting

    International Nuclear Information System (INIS)

    Hayakawa, H.; Kawamura, A.; Suto, O.; Kinoshita, Y.; Toda, Y.

    1985-01-01

    This paper describes the application of advanced computer technology to existing Japanese Boiling Water Reactor (BWR) nuclear power plants for backfitting. First we review the background of the backfitting and the objectives of backfitting. A feature of backfitting such as restrictions and constraints imposed by the existing equipment are discussed and how to overcome these restrictions by introduction of new technology such as highly efficient data transmission using multiplexing, and compact space saving computer systems are described. Role of the computer system in reliable NPS are described with a wide spectrum of TOSHIBA backfitting computer system application experiences. (author)

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  19. Reproduction of the flow-power map of the Laguna Verde power plant

    International Nuclear Information System (INIS)

    Amador G, R.; Gonzalez M, V.M.

    1993-01-01

    The National Commission of Nuclear Safety and Safeguards (CNSNS) requires to have calculation tools which allows it to make analysis independent of the behavior of the reactor core of Laguna Verde nuclear power plant (CNLV) with the purpose to support the evaluation and discharge activities of the fuel recharges licensing. The software package Fms (Fuel Management System) allows to carry out an analysis of the core of the BWR type reactors along the operation cycle to detect possible anomalies and/or helping in the fuel management. In this work it is reproduced the flow-power for the CNLV using the Presto code of the Fms software package. The comparison of results with the map used by the operators of CNLV shows good agreement between them. Another exercise carried out was the changes study that the axial and radial power outlines undergo as well as the thermohydraulic parameters (LHGR, APLHGR, CPR) when moving a control rod. The obtained results show that is had the experience to effect analysis of the reactor behavior using the Presto-Fms code therefore the study of the rest of the software package for the obtention of nuclear parameters used in this code is recommended. (Author)

  20. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.

    2009-01-01

    In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

  1. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

    2017-03-15

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  2. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    International Nuclear Information System (INIS)

    Badea, Aurelian F.; Cacuci, Dan G.

    2017-01-01

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  3. Advances in Nuclear Power Plant Water Chemistry in Reducing Radiation Exposure

    International Nuclear Information System (INIS)

    Febrianto

    2005-01-01

    Water quality in light water reactor in Pressurized Water Reactor as well as in Boiling Water Reactor has being gradually improved since the beginning, to reduce corrosion risk and radiation exposure level. Corrosion problem which occurred to both type of reactors can reduce the plants availability, increase the operation and maintenance cost and increase the radiation exposure. Corrosion and radiation exposure risk in both reactor rare different. BWR type reactor has more experiences in corrosion problem because at the type of reactor lets water to boil in the core, while at PWR type reactor, water is kept not to boil. The BWR reactor has also higher radiation exposure rather than the PWR one. Many collaborative efforts of plants manufacturers and plant operator utilities have been done to reduce the radiation exposure level and corrosion risk. (author)

  4. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  5. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  6. Application of ABWR construction database to nuclear power plant project

    International Nuclear Information System (INIS)

    Takashima, Atsushi; Katsube, Yasuhiko

    1999-01-01

    Tokyo Electric Power Company (TEPCO) completed the construction of Kashiwazaki-Kariwa Nuclear Power Station Unit No. 6 and No. 7 (K-6/7) as the first advanced boiling water reactors (ABWR) in the world successfully. K-6 and K-7 started their commercial operations in November, 1996 and in July, 1997 respectively. We consider ABWR as a standard BWR in the world as well as in Japan because ABWR is highly reputed. However, because the interval of our nuclear power plant construction is going to be longer, our engineering level on plant construction will be declining. Hence it is necessary for us to maintain our engineering level. In addition to this circumstance, we are planning to wide application of separated purchase orders for further cost reduction. Also there is an expectation for our contribution to ABWR plant constructions overseas. As facing these circumstances, we have developed a construction database based on our experience for ABWR construction. As the first step of developing the database for these use, we analyzed our own activities in the previous ABWR construction. Through this analysis, we could define activity units of which the project consists. As the second step, we clarified the data which are treated in each activity unit and the interface among them. By taking these steps, we could develop our database efficiently. (author)

  7. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  8. Validation of a methodology for the study of generation cost of electric power for nuclear power plants

    International Nuclear Information System (INIS)

    Ortega C, R.F.; Martin del Campo M, C.

    2004-01-01

    It was developed a model for the calculation of costs of electric generation of nuclear plants. The developed pattern was validated with the one used by the United States Council for Energy Awareness (USCEA) and the Electric Power Research Institute (EPRI), in studies of comparison of alternatives for electric generation of nuclear plants and fossil plants with base of gas and of coal in the United States described in the guides calls Technical Assessment Guides of EPRI. They are mentioned in qualitative form some changes in the technology of nucleo electric generation that could be included in the annual publication of Costs and Parameters of Reference for the Formulation of Projects of Investment in the Electric Sector of the Federal Commission of Electricity. These changes are in relation to the advances in the technology, in the licensing, in the construction and in the operation of the reactors called advanced as the A BWR built recently in Japan. (Author)

  9. Fukushima, two years later, modification requirements in nuclear power plants

    International Nuclear Information System (INIS)

    Sanchez J, J.; Camargo C, R.; Nunez C, A.; Mendoza F, J. E.; Salmeron V, J. A.

    2013-10-01

    The occurred events in the nuclear power plant of Fukushima Daiichi as consequence of the strong earthquake of 9 grades in the Richter scale and the later tsunami with waves estimated in more than 14 meters high began a series of important questions about the safety of the nuclear power plants in operation and of the new designs. Firstly, have allowed to be questioned on the magnitudes and consequences of the extreme external natural events; that can put in risk the integrity of the safety barriers of a nuclear power plant when being presented in a multiple way. As consequence of the events of the Fukushima Daiichi NPP, the countries with NPPs in operation and /or construction carried out evaluations about their safety operation. They have also realized evaluations about accidents and their impact in the safety, analysis and studies too that have forced to the regulatory bodies to continue a systematic and methodical revision of their procedures and regulations, to identify the possible improvements to the safety in response to the events happened in Japan; everything has taken it to determine the necessity to incorporate additional requirements to the nuclear power plants to mitigate events Beyond the Design Base. Due to Mexico has the nuclear power plant of Laguna Verde, with two units of BWR-5 type with contention Mark III, some the modifications can be applicable to these units to administrate and/or to mitigate the consequences of the possible occurrence of an accident Beyond the Design Base and that could generate a severe accident. In this work an exposition is presented on the modification requirements to confront external natural events Beyond the Design Base, and its application in our country. (Author)

  10. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

  11. 1100 MW BWR power station

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Now, the start-up test of No. 2 plant in Fukushima No. 2 Nuclear Power Station is smoothly in progress, and the start of its commercial operation is scheduled at the beginning of 1984. Here, the main features of No. 2 plant including piping design are explained. For No. 2 plant, many improving measures were adopted as the base plant of the improvement and standardization project of the Ministry of International Trade and Industry, such as the adoption of Mark-2 improved PCV, the adoption of an intermediate loop in the auxiliary cooling system, one-body forging of the lower end cover of the reactor pressure vessel, the adoption of many curved pipes, the adoption of large one-body structural components in reactor recirculation system piping and so on, which are related to the reduction of radiation exposure and the improvement of plant reliability in operation and regular inspection. Also, in order to do general adjustment in the arrangement of equipment and piping and in route design, and to establish the rational construction work plan, model engineering was adopted. In No. 2 plant, a remote-controlled automatic and semiautomatic ultrasonic flaw detection system was adopted to reduce radiation exposure in in-service inspection. Automatic welding was adopted to improve the quality. (Kako, I.)

  12. BWR startup and shutdown activity transport control

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

    2010-07-01

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

  13. Intermediate size LWR plant study for process heat plus power: main report. Volume 2 of 3 volumes

    International Nuclear Information System (INIS)

    Head, M.A.

    1977-01-01

    The study of process heat-plus-power generation from 900 to 1200 MWt BWRs was initiated by analysis of the smaller BWR plants, which have substantial operating experience. The Muhleberg 306 MWe BWR (four years of operation and 74% capacity factor) and the Humboldt Bay, natural circulation system (ten years of operation and 69% capacity factor) were identified as leading plants. The Muhleberg design, cost, and operating data therefore were analyzed in further detail. Analysis was also conducted on the Humboldt Bay unit, even though its 63 MHe size is not directly applicable. In this case, attention was focused on the potential for updating the natural circulation concept to the 1000 MWt level. Cost analyses were conducted on actual costs incurred, including correction indices for escalation to 1976. The changes in licensing requirements were evaluated and their cost impact estimated. Some indication of the economies possible through standardization and use of modular construction were estimated by reference to earlier work under the nuclear park (or energy center) study program. Consideration was directed to the organizational needs for design, construction, and project management, if an effective program were to be undertaken. The importance of a demonstration project as a means of establishing a design standard for duplication or replication (and the program requirements to bring about such a demonstration project) were then addressed. Finally, capital costs developed from the actual plant costs were used to determine power generation costs

  14. Safety evaluation of BWR off-gas treatment systems

    International Nuclear Information System (INIS)

    Schultz, R.J.; Schmitt, R.C.

    1975-01-01

    Some of the results of a safety evaluation performed on current generic types of BWR off-gas treatment systems including cooled and ambient temperature adsorber beds and cryogenics are presented. The evaluation covered the four generic types of off-gas systems and the systems of five major vendors. This study was part of original work performed under AEC contract for the Directorate of Regulatory Standards. The analysis techniques employed for the safety evaluation of these systems include: Fault Tree Analysis; FMECA (Failure Mode Effects and Criticality Analysis); general system comparisons, contaminant, system control, and design adequacy evaluations; and resultant Off-Site Dose Calculations. The salient areas presented are some of the potential problem areas, the approach that industry has taken to mitigate or design against potential upset conditions, and areas where possible deficiencies still exist. Potential problem areas discussed include hydrogen detonation, hydrogen release to equipment areas, operator/automatic control interface, and needed engineering evaluation to insure safe system operation. Of the systems reviewed, most were in the category of advanced or improved over that commonly in use today, and a conclusion from the study was that these systems offer excellent potential for noble gas control for BWR power plants where more stringent controls may be specified -- now or in the future. (U.S.)

  15. Operating experience with BWR nuclear power

    International Nuclear Information System (INIS)

    Bonsdorf, Magnus von.

    1986-01-01

    The two-unit nuclear power station in Olkiluoto on the western coast of Finland produces about 20 per cent of the electricity consumption of the country. The first unit, TVO-I was first connected to the national grid in September 1978 and TVO-II in February 1980. The original rated power output of each unit was 660 MWe, corresponding to the thermal power of 2000 MW from the reactor. Technical modifications allowed the power to be uprated by 8%. The operating statistics (load factors etc.) are given and the outage experience discussed. The radiological history shows very low radioactivity and dose levels have been maintained at the plant. (UK)

  16. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  17. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  18. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  19. Reliability and improvement of RODOS results for a BWR plant; Erhoehung der Zuverlaessigkeit der RODOS-Ergebnisse fuer eine SWR-Anlage

    Energy Technology Data Exchange (ETDEWEB)

    Loeffler, H.; Cester, F.; Sonnenkalb, M.; Klein-Hessling, W.; Voggenberger, T.

    2009-06-15

    Decision support systems such as RODOS aim to support the responsible authorities by providing estimates for the possible radiological consequences in case of an event in a nuclear plant. The prognosis of quantity, composition and time of occurrence of a release from the plant (''source term'') in the so-called pre-release phase is one of the foundations with high relevance for this purpose. Within previous projects source term prognosis tools have been developed and applied exemplarily for a PWR. At the end of 2005 GRS has finalized a PSA level 2 for a plant of the SWR-69 type. On this basis improved versions of the source term prognosis tools QPRO (probabilistic) and ASTRID (deterministic) have been created for a BWR and tested in an emergency exercise in a BWR. The further development of QPRO has been related in particular to the structure of the probabilistic network and the precalculated source terms. The activities for the adaptation of ASTRID focus on the creation of the dataset for the BWR coolant loop and the containment. In the emergency exercise the manageability of QPRO but also of ASTRID has been proven. Further, the first phases of the accident progression have been well identified. However, the exercise scenario developed into a very unlikely sequence with partial core melt, and the reactor building ventilation was shut off just at a critical moment. Therefore the source term prognoses deviate from the exercise scenario. Starting from these experiences with the development and application of QPRO and ASTRID recommendations are given for the further improvement of the reliability of the source term prognosis for RODOS. In general it can be stated that the development status of QPRO and ASTRID is definitely advanced compared to the presently still prevailing source term prognosis methods. Therefore it is recommended to develop plant specific versions of these codes and to apply them.

  20. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  1. Large bundle BWR test CORA-18: Test results

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

    1998-04-01

    The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

  2. Optimization and improvement of the technical specifications for Santa Maria de Garona and Cofrentes nuclear power plants; Optimizacion y mejora de las especificaciones tecnicas de funcionamiento para las centrales de Santa Maria de Garona y Cofrentes

    Energy Technology Data Exchange (ETDEWEB)

    Norte Gomez, M D [Empresarios Agrupados, A.I.E., Madrid (Spain); Alcantud, F [Iberdrola C.N., Cofrentes, (Spain); Hoyo, C del [Nuclenor C.N., Santa Maria de Garona (Spain)

    1993-12-15

    Technical Specifications (TS) form one of the basic documents necessary for licensing nuclear power plants and are required by the Government in accordance with Article 26 of the Regulation for Nuclear and Radioactive Facilities. They contain specific plant characteristics and operating limits to provide adequate protection for the safety and health of operators and the general public. For operator actuation, TS include all the surveillance requirements and limiting operating conditions (operation at full power, startup, hot and cold shutdown, and refueling outage) of safety-related systems. They also include the conventional support systems which are necessary to keep the plant in a safe operating conditioner to bring it to safe shutdown in the event of incidents or hypothetical accidents. Because of the large volume of information contained in the TS, the NRC and American utility owners began to simplify and improve the initial standard TS, which has given way to the development of a TS Optimization Program in the USA under the auspices of the NRC. Empresarios Agrupados has been contracted by the BWR Spanish Owners' Group (GPE-BWR) to develop optimized TS for the Santa Maria de Garona and Cofrentes Nuclear Power Plants. The optimized and improved TS are simplified versions of the current ones and facilitate the work of plant operators. They help to prevent risks, and reduce the number of potential transients caused by the large number of tests required by current TS. Plant operational safety is enhanced and higher effective operation is achieved. The GPE-BWR has submitted the first part of the optimized TS with their corresponding Bases to the Spanish Nuclear Council (CSN), for comment and subsequent approval. Once the TS are approved by the Spanish Nuclear Council, the operators of the Santa Maria de Garona and Cofrentes Nuclear Power Plants will be given a training and adaptation course prior to their implementation. (author)

  3. State of the art of second international exercise on benchmarks in BWR reactors

    International Nuclear Information System (INIS)

    Verdu, G.; Munoz-Cobo, J. L.; Palomo, M. J.; Escriva, A.; Ginestar, D.

    1998-01-01

    This is a second in series of Benchmarks based on data from operating Swedish BWRs. The first one concerned measurements made in cycles 14,15 16 and 17 at Ringhals 1 Nuclear Power Plant and addressed predictive power of analytical tools used in BWR stability analysis. Part of the data was disclosed only after participants had provided their results. This work has been published in the report: NEA/NSC/DOC(96)22, November 1996. In this report it was recognised that there is a need for better qualification of the applied noise analysis methods. A follow up Benchmark was thus proposed dedicated to the analysis of time series data and including the evaluation of both global and regional stability of Forsmarks 1 and 2 Nuclear Power Plant. In this second Benchmark have participated Forsmarks Kraftgrupp AB,NEA Nuclear Science Committee, CSN Consejo de Seguridad Nuclear and Department of Chemical and Nuclear Engineering of Polytechnic University of Valencia. (Author)

  4. Comparative of fuel cycle cost for light water nuclear power plants; Uporedna analiza cene gorivnog ciklusa lakovodne nuklearne elektrane

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Dimitrijevic, Z [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1978-07-01

    Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

  5. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

  6. Statistical analysis about corrosion in nuclear power plants

    International Nuclear Information System (INIS)

    Naquid G, C.; Medina F, A.; Zamora R, L.

    1999-01-01

    Nowadays, it has been carried out the investigations related with the structure degradation mechanisms, systems or and components in the nuclear power plants, since a lot of the involved processes are the responsible of the reliability of these ones, of the integrity of their components, of the safety aspects and others. This work presents the statistics of the studies related with materials corrosion in its wide variety and specific mechanisms. These exist at world level in the PWR, BWR, and WWER reactors, analysing the AIRS (Advanced Incident Reporting System) during the period between 1993-1998 in the two first plants in during the period between 1982-1995 for the WWER. The factors identification allows characterize them as those which apply, they are what have happen by the presence of some corrosion mechanism. Those which not apply, these are due to incidental by natural factors, mechanical failures and human errors. Finally, the total number of cases analysed, they correspond to the total cases which apply and not apply. (Author)

  7. Chemistry in power plants 2011; Chemie im Kraftwerk 2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    Within the VGB Powertech conference from 25th to 27th October, 2011, in Munich (Federal Republic of Germany), the following lectures and poster contributions were presented: (1) The revised VGB standard for water-steam-cycle Chemistry; (2) Switchover from neutral operation to oxygen treatment at the power station Stuttgart-Muenster of EnBW Kraftwerke AG; (3) Steam contamination with degradation products of organic matters present in the feedwater of the Lanxess-Rubber cogeneration plant; (4) Laboratory scale on-line noble metal deposition experiments simulating BWR plant conditions; (5) Building a new demin installation for the power plant EPZ in Borssele; (6) Replacement of the cooling tower installations in the nuclear power plant Goesgen-Daenien AG; (7) Aging of IEX resins in demin plants - Cost optimisation by adaptation of regenerants; (8) The largest DOW trademark EDI System at a combined cycled plant in Europe; (9) Upgrading river Main water to boiler feed water - Experiences with ultrafiltration; (10) Experiences with treatment of the water-steam-cycle in the RDF power plant Nehlsen Stavenhagen with film-forming amines; (11) Comparative modelling of the bubbles thermal collapse and cavitations for estimation of bubbles collapse influence; (12) Overcoming the steam quality - issues from an HRSG for the production of process steam; (13) Legionella - new requirements for power plant operation; (14) How the right chemistry in the FGD helps to improve the removal in the waste water treatment plant; (15) High efficiency filtration in dry/semi-dry FGD plants; (16) Expanding the variety of renewable fuels in the biomass power plant Timelkam using the chemical input control; (17) Corrosion, operating experiences and process improvements to increase the availability and operating time of the biomass power plant Timelkam; (18) The influence of temperature on the measurement of the conductivity of highly diluted solutions; (19) A multiparameter instrumentation approach

  8. Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

    Full Text Available The loss-of-coolant accident (LOCA simulation in the boiling water reactor (BWR of Laguna Verde Nuclear Power Plant (LVNPP at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

  9. Evaluation Of The Exclusion And Low Population Areas Around A Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tawfik, F.S.

    2011-01-01

    Being adjacent to the nuclear power plant (NPP) the exclusion area (EA) is the area of greatest importance. It essentially defines a buffer zone where the public has no access. It helps to define the fenced plant area, the site area and the public area. Also, the low population area is the area immediately surrounding the exclusion area near a licensed reactor in terms of public safety and the ability of residents to get away from the plant in an emergency. This study clarifies their significance and reviews the international approach on them. Assuming the nuclear power plant site at the north coast of Egypt, the exclusion area and low population area are determined according to CFR (2002). In this method, a maximum possible amount of radioactivity release (called a source term) should be assumed. The boiling water reactor (BWR) with a power 1000 MW was used to carry the calculation and assuming a severe loss of coolant accident with meltdown of reactor. The site specific data have been collected, investigated and processed. The effect of the degree of atmospheric stability and building width of the plant were examined. The proceeding factors that control the determination of exclusion area and low population area should be taken into consideration in the site evaluation stage and design basis of NPP to set a minimum distances for them

  10. Upgrading of fire safety in Indian nuclear power plants

    International Nuclear Information System (INIS)

    Agarwal, N.K.

    1998-01-01

    Indian nuclear power programme started with the installation of 2 nos. of Boiling Water Reactor (BWR) at Tarapur (TAPS I and II) of 210 MWe each commissioned in the year 1996. The Pressurized Heavy Water Reactor (PHWR) programme in the country started with the installation of 2x220 MWe stations at Rawatbhatta near Kota (RAPS I and II) in the State of Rajasthan in the sixties. At the present moment, the country has 10 stations in operation. Construction is going on for 4 more units of 220 MWe where as work on two more 500 MWe units is going to start soon. Fire safety systems for the earlier units were engineered as per the state-of-art knowledge available then. The need for review of fire protection systems in the Indian nuclear power plants has also been felt since long almost after Brown's Ferry fire in 1975 itself. Task forces consisting of fire experts, systems design engineers, O and M personnel as well as the Fire Protection engineers at the plant were constituted for each plant to review the existing fire safety provisions in details and highlight the upgradation needed for meeting the latest requirements as per the national as well as international practices. The recommendations made by three such task forces for the three plants are proposed to be reviewed in this paper. The paper also highlights the recommendations to be implemented immediately as well as on long-term basis over a period of time

  11. Development of neural network for analysis of local power distributions in BWR fuel bundles

    International Nuclear Information System (INIS)

    Tanabe, Akira; Yamamoto, Toru; Shinfuku, Kimihiro; Nakamae, Takuji.

    1993-01-01

    A neural network model has been developed to learn the local power distributions in a BWR fuel bundle. A two layers neural network with total 128 elements is used for this model. The neural network learns 33 cases of local power peaking factors of fuel rods with given enrichment distribution as the teacher signals, which were calculated by a fuel bundle nuclear analysis code based on precise physical models. This neural network model studied well the teacher signals within 1 % error. It is also able to calculate the local power distributions within several % error for the different enrichment distributions from the teacher signals when the average enrichment is close to 2 %. This neural network is simple and the computing speed of this model is 300 times faster than that of the precise nuclear analysis code. This model was applied to survey the enrichment distribution to meet a target local power distribution in a fuel bundle, and the enrichment distribution with flat power shape are obtained within short computing time. (author)

  12. Two approaches to meeting the economic challenge for advanced BWR designs

    International Nuclear Information System (INIS)

    Rao, A.S.; Sawyer, C.D.

    1997-01-01

    This paper presents the design overview and approach to addressing the aforementioned economic challenges for two Advanced BWR designs. The first plant is the ABWR and the second is the ESBWR. The ABWR relies on proven technology and components and an extensive infrastructure that has been built up over the last 20 years. Because it has proven and standards safety systems it has very limited uncertainty regarding licensing. Finally, it relies on the economies of scale and overall design flexibility to improve the overall economics of power generation. The ESBWR on the other hand has taken an innovative approach to reduce systems and components to simplify the overall plant to improve plant economics. The overall plant design is indeed simpler, but improved economics required reliance on some economies of scale also. This design embodied in the ESBWR, also has minimized the overall development cost by utilizing features and components from the ABWR and SBWR technology programs

  13. 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

  14. An economic analysis of BWR control rod blade management strategies. Final report

    International Nuclear Information System (INIS)

    Welsh, J.

    1995-12-01

    Nuclear power plants have available a number of alternative courses of action that can contribute to the reduction of personnel exposure to radiation. Possible actions at boiling water reactor (BWR) plants include accelerating the replacement of high-cobalt control rod blades (CRB) or the blades' high-cobalt pins and rollers with low or non-cobalt substitutes. To help utilities understand the exposure reduction and the economic costs and benefits associated with management alternatives, such as accelerated replacement of blades, pins and rollers, EPRI has initiated a project called Cost/Benefit Software for Analyses of Radiation Control Measures (RP1935-32). Through this project EPRI will incorporate engineering-economic techniques into a series of analytical tools that will provide useful insights about alternative exposure reduction options. Prototype software has been developed in an Excel worksheet to analyze issues associated with BWR control rod blade management options. The CRB replacement problem framework and analysis methodology incorporated into the software tool will help plant managers consider explicitly key engineering and economic issues that are relevant to exposure reduction decisions. This tool generates results that can help plant managers make decisions that are fiscally wise by showing all the cost and benefit implications associated with a management action under consideration. This report describes the general analytical approach for evaluating exposure reduction alternatives. The methodology used to analyze blade and pin and roller replacement alternatives, and the results of a case study application of the methodology and the software prototype at Commonwealth Edison

  15. Technology and costs for decommissioning the Swedish nuclear power plants

    International Nuclear Information System (INIS)

    1986-05-01

    The study shows that, from the viewpoint of radiological safety, a nuclear power plant can be dismantled immediately after it has been shut down and the fuel has been removed, which is estimated to take about one year. Most of the equipment that will be used in decommissioning is already available and is used routinely in maintenance and rebuilding work at the nuclear power plants. Special equipment need only be developed for dismantlement of the reactor vessel and for demolishing of heavy concrete structures. The dismantling of a nuclear power plant can be accomplished in about five years, with an average labour force of about 200 men. The maximum labour force required for Ringhals 1 has been estimated at about 500 men during the first years, when active systems are being dismantled in a number of fronts in the plant. During the last years when the buildings are being demolished, approximately 50 men are required. In order to limit the labour requirement and the dose burden to the personnel, the material is taken out in as large pieces as possible. The cost of decommissioning a boiling water reactor (BWR) of the size of Ringhals 1 has been estimated to be about MSEK 540 in January 1986 prices, and for a pressurized water reactor (PWR, Ringhals 2) about MSEK 460. The cost for the other Swedish nuclear power plants lie in the range of MSEK 410-760. These are the direct cost for the decommissioning work, to which must be added the costs of transportation and disposal of the decommissioning waste, about 100 000 m/sup3/. These costs have been estimated to be about MSEK 600 for the 12 Swedish reactors. (author)

  16. Occupational radiation exposure of personnel in German light-water type nuclear power plant in the year 1985

    International Nuclear Information System (INIS)

    Urbahn, H.; Mueller, W.; Marx, H.

    1988-01-01

    1985 was a good year with regard to radiation protection for the German nuclear power stations with BWR and PWR. There were favourable values for the collective dose. There was an increase in the collective dose only in some power plants because of dose-intensive repair, replacement and retrofit work. The average personal doses of the internal staff as well as of the external staff showed quite favourable values also in comparison to the previous years. The increased average personal doses occurred generally in the older plants. This is caused, on the one hand, by dose-intensive repair, replacement and retrofit work, on the other hand, it reflects the different deployment of labour. The dose values for the job doses of the inspection and maintenance work were within the range of the empirical and expected values. High dose values occurred in general only in retrofit measures, which were carried out namely because of obligations by the regulatory authority. Only a small proportion of the collective dose was due to repair work. In the case of the BWR, the dose values for the job doses of fuel element replacement, maintenance and inspection work were within the range of the empirical and expected values. Retrofit and repair work did not contribute much to the collective dose. (orig./HP) [de

  17. Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.; Nuenighoff, K.; Allelein, H.J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energie- und Klimaforschung (IEK), Sicherheitsforschung und Reaktortechnik (IEK-6)

    2011-07-01

    Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

  18. Applicability study on a ceramic filter with hot-test conducted in a BWR plant

    International Nuclear Information System (INIS)

    Yamada, K.; Shirai, T.; Wada, M.; Nakamizo, H.

    1991-01-01

    Radioactive crud removal and filtration performance recovery by backwashing were examined with a BWR plant pool water using a ceramic filter element, 0.1 micron in nominal pore size and 0.2m 2 in filtration area. Totally 1114 hours filter operation were accumulated. Ten backwashings were accomplished during the test period. The following results were obtained. (1) Radioactive crud concentration in the filter effluent remained below 10 5 Bq/m 3 . (2) Both pressure loss through the filter and dose rate at the filter vessel surface were recovered to the initial level by each backwashing. The surface dose rate after backwashing was approximately 0.01mSv/h. According to these test results, it is confirmed that the ceramic filter is appropriate for the treatment of highly crud concentrated radioactive liquid, which is generated in nuclear facilities, such as spent fuel reprocessing plants. (author)

  19. Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

    2011-07-01

    Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

  20. Analysis of the noise of the jet pumps of the Unit 2 of the Laguna Verde nuclear power plant

    International Nuclear Information System (INIS)

    Castillo D, R.; Ortiz V, J.; Ruiz E, J.A.; Calleros M, G.

    2004-01-01

    The use of the analysis of noise for the detection of badly functioning of the components of a BWR it is a powerful tool in the determination of abnormal conditions of operation, during the life of a nuclear plant of power. From the eighties, some nuclear reactors have presented problems related with the jet pumps and the knots of the recirculation. The Regulatory Commission of the United States, in the I E bulletin 80-07, recommended to carry out a periodic supervision of the pressure drop of the jet pumps, to prevent structural failures. In this work, methods of analysis of noise are used for the detection of abnormal conditions of operation of the jet pumps of a BWR. Signals are analysed to low and high frequency of pressure drop with the NOISE software that is in development. The obtained results show the behavior of the jet pumps of jet 6 and 11 before and after a partial blockade in their throats where the pump 6 return to their condition of previous operation and the pump 11 present a new fall of pressure, inside the limit them permissible of operation. The methodology of the analysis of noise demonstrated to be an useful tool for the badly functioning detection, and you could apply to create a database to supervise the dynamic behavior of the jet pumps of an BWR. (Author)

  1. Prediction of droplet deposition around BWR fuel spacer by FEM flow analysis

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shinichi

    1997-01-01

    The critical power of the BWR fuel assembly has been remarkably increased. That increase mainly depends on the improvement of the spacer which keeps fixed gaps between fuel rods. So far, these improvements have been carried out on the basis of what developers consider to be appropriate and the results of mockup tests of the BWR fuel assembly. However, continued reliance on these approaches for the development of a higher performance fuel assembly will prove time-consuming and costly. Therefore, it is hoped that the spacer effects for the critical power can be investigated by computer simulation, and it is significantly important to develop the critical power prediction method. Direct calculation of the two-phase flow in a BWR fuel channel s still difficult. Accordingly, a new method for predicting the critical power was proposed. Our method consists of CFD (computer fluid dynamics) code based on the single-phase flow analysis method and the subchannel analysis code. To verify our method, the critical power predictions for various spacer geometries were performed. The predicted results of the critical power were compared with the experimental data. The result of the comparison showed a good agreement and the applicability of our method for various spacer geometries. (author)

  2. Ranking of input parameters importance for BWR stability based on Ringhals-1

    International Nuclear Information System (INIS)

    Gajev, Ivan; Kozlowski, Tomasz; Xu, Yunlin; Downar, Thomas

    2011-01-01

    Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Uncertainty calculations for BWR stability, based on the Wilks' formula, have been already done for the Ringhals-1 benchmark. In this work, these calculations have been used to identify and rank the most important parameters affecting the stability of the Ringhals-1 plant. The ranking has been done in two different ways and a comparison of these two methods has been demonstrated. Results show that the methods provide different, but meaningful evaluations of the ranking. (author)

  3. Modelling and use of the STUDS nuclear power plant simulator

    International Nuclear Information System (INIS)

    Blomberg, P.E.; Espefaelt, R.; Josefsson, R.; Schuch, N.

    1979-02-01

    The simulator models, belonging to the STUDS-family, which have been developed at Studsvik in cooperation with the Swedish utilities, are briefly described. The scope of the simulation is presented and the fundamental equations used are indicated. Different needs have led to a number of STUDS-versions for BWR and PWR type plants, primarily intended for application in the following fields: 1) transient analysis, 2) system design verification, 3) control system development, 4) testing of new on-line techniques for disturbance analysis, noise analysis, man-machine communication, etc, 5) training of power plant operators, 6) operational planning. The simulator was initially implemented on a hybrid computer system but more recent work has led to pure digital simulations maintaining the real time feature and adding features like snapshot and backtrack. The latest version for PWR is used at the Halden Project and in the general purpose COMPACT SIMULATOR: developed at Studsvik and made commercially available. (author)

  4. Diffusion bonded matrix of HGMF applied for BWR condensate water purification

    International Nuclear Information System (INIS)

    Soda, Fumitaka; Yukawa, Takao; Ito, Kazuyuki.

    1984-01-01

    High Gradient Magnetic Filter (HGMF) applied to the purification of power plant primary water has recently attracted much attention. In the application of HGMF to the water treatment of power plants, especially nuclear power plants, reliabillties of matrix (filtering medium) as well as removal performance for cruds (insoluble corrosion products) are considered to be important factors. To satisfy these factors, a new filtering medium named Diffision Bonded Matrix (DBM) has been developed and the test results are reported. Filtering efficiency and mechanical stiffness of DBM were examined using HGMF pilot test units consisting of 160 mm diameters x 240 mm length filter. The filtering velocity and the magnetic flux density used in this test were 800 m/h 5 kG, respectively. The filtering efficiencies and of 85-100% were obtained for artificial cruds for DBM. The DBM indicated slightly better filtering efficiency than for conventional wool matrix under the same filtering and matrix conditions. The DBM kept its original mechanical properties and very few pieces of fibers were broken off while the conventional wool matrix lost its volume elasticities and the considerable amount of fibers was broken off during the test operation. The results described here demonstrated the applicability of DBM for treatment of BWR primary water by High Gradient Magnetic Filter. (author)

  5. Metallurgical factors that contribute to cracking in BWR piping

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

  6. Kinetic analyses on startup and shutdown chemistry of BWR plant

    International Nuclear Information System (INIS)

    Domae, Masafumi; Fujiwara, Kazutoshi; Inagaki, Hiromitsu

    2012-09-01

    During startup and shutdown of Boiling Water Reactor (BWR) plants, temperature and dissolved oxygen (DO) concentration of reactor water change in a wide range. The changes result in variation of conductivity and pH of the reactor water. It has been speculated that the water chemistry change is due to dissolution of the oxides on fuel claddings and structural materials. However, detailed mechanism is not known. In the present paper, trend of recent water chemistry in several BWR plants during startup and shutdown is presented. Conductivity and pH are convenient indication of coolant purity. We tried to clarify the mechanism of the change in the conductivity and the pH value during startup and shutdown, based on the water chemistry data measured. In the water chemistry data, change in chromate concentration and Ni 2+ concentration is rather large. It is assumed that change in the chromate concentration and the Ni 2+ concentration results in the time variation of the conductivity and the pH value. It is reasonable to consider that the increase in the chromate concentration and the Ni 2+ concentration is ascribed to dissolution of Cr oxides and Ni oxides, respectively. A model of dissolution of the Cr oxides and the Ni oxides is proposed. A concept of finite inventory of the Cr oxides and the Ni oxides in the coolant system is introduced. The model is as follows. Chromate is generated by oxidation of the Cr oxides and the Cr dissolution rate depends on the DO concentration. The dissolution rate of chromate is in proportion to DO concentration, the inventory of Cr and difference between solubility limit and the chromate concentration. On the other hand, Ni 2+ is formed by dissolution of the Ni oxides, and DO is not necessary in this process. The dissolution rate of Ni 2+ is in proportion to the inventory of Ni and difference between solubility limit and the Ni 2+ concentration. Coolant is continuously purified, and the chromate concentration and the Ni 2+ concentration

  7. Cost benefit analysis of instrumentation, supervision and control systems for nuclear power plants

    International Nuclear Information System (INIS)

    Hagen, P.

    1973-08-01

    A cost benefit analysis is carried out on a BWR type reactor power plant in which an on-line computer performs plant supervision, reporting, logging, calibration and control functions, using display devices and plotters, while an off-line computer is available for bigger jobs such as fuel management calculations. All on-line functions are briefly described and specified. Three types of computer system are considered, a simplex system, a dual computer system and a multi-processor system. These systems are analysed with respect to reliability, back-up instrumentation requirements and costs. While the multiprocessor system gave in all cases the lowest annual failure costs, the margin to the duplex system was so small that hardware, maintenance and software costs would play an important role in making a decision. (JIW)

  8. 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

  9. PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

    OpenAIRE

    MELARA SAN ROMÁN, JOSÉ

    2016-01-01

    [EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

  10. Management of radioactive waste from a major core damage in a BWR power plant

    International Nuclear Information System (INIS)

    Elkert, J.; Christensen, H.; Torstenfelt, B.

    1990-01-01

    Large amounts of fission products would be released in case of a major core damage in a nuclear power reactor. In this theoretical study the core damage is caused by a loss of coolant accident followed by a complete loss of all electric power for about 30 minutes resulting in the release of 10% of the core inventory of noble gases. A second case has also been briefly studied, in which the corresponding core damage is supposed to be created merely by the complete loss of electric power during a limited time period. It appears from the study that the radioactive waste generated as a consequence of an accident of the extent can be managed in the reference reactor with only minor modifications required in the waste plant. The detailed results of the study are reactor specific, but many of the findings and recommendations are generally applicable. (author) 28 refs

  11. Analysis of risk-dominant sequences by MAAP3.0 for Kuosheng Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lin, J.D.; Chieng, C.C.; Wang, T.K.; Hsiue, R.K.

    1987-01-01

    Kuosheng Nuclear Power Plant is the first operating model-6/Mark III boiling water reactor (BWR6/Mark III) in the world, and a probabilistic risk assessment (PRA) has been performed to determine the likely frequencies of core melt accidents and the magnitude, composition, and fraction of fission products released in these accidents. The final report of this PRA indicates that categories 8 and 15 are ranked No. 1 by risk index (the product of release frequency and release fraction) and release frequency, respectively. The dominant contributors of these two categories are frequent earthquakes and typhoons

  12. Radwaste reduction through use of condensate non-precoat filters at Perry Nuclear Power Plant

    International Nuclear Information System (INIS)

    Hartman, J.L.

    1996-01-01

    The concept of non-precoat filtration to remove iron oxide without generating traditional resin and fiber precoat material, which is expensive to dispose of as radwaste, is rapidly becoming a trend of the 90's. In 1991, in response to escalating radwaste disposal costs, Perry Nuclear Power Plant was the first BWR to install non-precoat filters (septa) on a full scale basis in a Condensate System. To date, non-precoat septa from three vendors; including Memtec Power Generation have been installed and operated at Perry. It is important that other utilities considering this technology be given as much open-quotes real lifeclose quotes operational data as possible. This paper presents the data from three different companies and provides recommendations for consideration

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

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

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

  14. List of key words with classification for a standard safety report for nuclear power plants with PWR or BWR

    International Nuclear Information System (INIS)

    1976-01-01

    Under the efforts of improving the licensing procedure for nuclear power plants, the Federal Minister of the Interior set up a task group of experts of the manufacturers and operators of nuclear power plants, the assessors (Technische Ueberwachungsvereine, TUeVs), the Institute for Reactor Safety of the TUeVs, the licensing authorities of the Laender, and the Federal Ministry of the Interior which worked out a list of key words for writing the safety report for nuclear power plants with PWRs and BWRs. This list of key words is published herewith in order to encourage its application when writing or assessing safety reports for nuclear power plants and in order to present the opportunity to make proposals for improvement to a group as large as possible. At a later date, it is intended to incorporate the list of key words as soon as sufficient experience from the practical application will justify this, it is intended to incorporate the list of key words in a general administrative regulation. (orig.) [de

  15. Comparison of BWR-6 pressurization transients with one-dimensional and point kinetics

    International Nuclear Information System (INIS)

    Serra, J.M.; Mata, P.; Cronin, J.T.

    1992-01-01

    This paper focuses on the differences between the results of core reload licensing calculations for the BWR-6 plant when performed with a one-dimensional (1-D) versus a point kinetics model. More specifically, the improvement in critical power ratio which would be expected from a change in methods from a point to a 1-D kinetics core wide transient calculation for pressurization transients is investigated. To qualitatively assess critical power ratio (CPR) improvement, core wide transient and hot channel calculations of a generator load rejection with failure of the steam by-pass system and a feedwater controller failure of maximum demand are performed with both, point and 1-D kinetics models in the core wide simulation. Additionally, a sensitivity study on the frequency of power shape function updating in the 1-D kinetics calculation is performed

  16. Way to LWR nuclear power stations taking root

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1994-01-01

    In this report, as for BWR nuclear power stations, the countermeasures to initial troubles, the circumstances of making bold start for the improvement and standardization after the remarkable lowering of the rate of operation, the development of A-BWR aiming at the most excellent BWR plant and the effort to reduce the construction cost are looked back, and the opinion on advancing nuclear energy development hereafter is mentioned. As the initial troubles, the thermal fatigue of reactor feedwater nozzles, the cracking of return water nozzles in control rod drive, the damage of fuel elements and others occurred. The circumstances till starting the project of improvement and standardization, and the investigation by the ad hoc committee and the results are reported. The feasibility study on A-BWR was carried out from July, 1978 to October, 1979. The basic design and the research and development for verification and the optimization design were carried out till December, 1985, as the result, the improvement of the plant performance was obtained. The effort for reducing the construction cost was successful. (K.I.)

  17. Isolation valve control device for nuclear power plant

    International Nuclear Information System (INIS)

    Yukinori, Shigeru.

    1990-01-01

    The present invention provides an isolation valve control device for detecting pipeline rupture accidents in a BWR type nuclear power plant at an early stage to close an isolation valve thereby reducing the amout of radioactivity released to the circumstance. That is, isolation valves are disposed in the pipeline for each of the systems in the nuclear power plant and flow ratemeters are disposed to at least two positions in each of the pipelines. If a meaningful difference is shown for the measured values by these flow ratemeters, the isolation valve is closed. In this way, if pipeline rupture such as leak before break (LBB) is caused to a portion of a system pipelines, the measured value from the flow ratemeters at the downstream of the pipeline is lowered. Accordingly, when a meaningful difference is formed between the value of the flow ratematers at the upstream and the downstream, occurrence of pipe rutpture between both of the flow ratemeters can be detected. As a result, the isolation valves of the system can be closed. According to the present invention, it is possible to detect the pipeline rupture at an early stage irrespective of the kind of the systems, diameter of the pipelines and the magnitude of the ruptured area, and the isolation valve can be closed. (I.S.)

  18. Contributions of welding technology to power plant performance

    International Nuclear Information System (INIS)

    Childs, W.J.

    1995-01-01

    Welding repairs can be a very major factor in the time and cost of maintenance outages of a power plant. The use of advanced equipment and procedures for welding can contribute significantly to reducing maintenance costs and increasing reliability. Plant failures have too often been caused by problems associated with welding, some due to improper choice of base materials, others due to welding defects. For example, stress corrosion cracking in weldments in BWR austenitic stainless steel piping was a major source of loss of availability in the 1980s. Examples of the use of improved welding equipment and procedures has been demonstrated to reduce outage time and improved weld integrity in several major areas. New welding techniques, such as laser welding, have the potential for addressing maintenance problems that can not be addressed at all with conventional welding technology and/or may provide a means of reducing greatly the time and cost of welding fabrications or repair. Methods of ensuring that the best available technology is applied in weld repair is a major problem today in the utility industry. Solutions need to be sought to remedy this situation. The key role of welding in minimizing plant outages is being recognized and steps taken to further the development and use of optimum technology

  19. Risk-informed decision-making analysis for the electrical raceway fire barrier systems on a BWR-4 plant

    International Nuclear Information System (INIS)

    Wu, Ching-Hui; Lin, Tsu-Jen; Kao, Tsu-Mu; Chen, Chyn-Rong

    2003-01-01

    This paper describes a risk-informed decision-making approach used to resolve the fire barrier issue in a BWR-4 nuclear plant where Appendix R separation requirements cannot be met without installing additional fire protection features such as electrical raceway fire barrier system. The related risk measures in CDF (core damage frequency) and LERF (large early release frequency) of the fire barrier issue can be determined by calculating the difference in plant risks between various alternative cases and that met the requirement of the Appendix R. In some alternative cases, additional early-detection and fast-response fire suppression systems are suggested. In some other cases, cable re-routing of some improper layout of non-safety related cables are required. Sets of fire scenarios are re-evaluated more detailed by reviewing the cable damage impact for the BWR-4 plant. The fire hazard model, COMPBRM III-e, is used in this study and the dominant results in risk measures are benchmarked with the CFD code, FDS 2.0, to ensure that the risk impact of fire barrier is estimated accurately in the risk-informed decision making. The traditional deterministic qualitative methods, such as defense-in-depth, safety margin and post-fire safety shutdown capability are also proceeded. The value-impact analysis for proposed alternatives of fire wrapping required by Appendix R has been completed for technical basis of the exemption on Appendix R application. The outcome of the above analysis should be in compliance with the regulatory guidelines (RG) 1.174 and 1.189 for the applications in the risk-informed decision-making of the fire wrapping issues. (author)

  20. Validation of the Engineering Plant Analyzer methodology with Peach Bottom 2 stability tests

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Mallen, A.N.; Cheng, H.S.; Wulff, W.

    1994-01-01

    The Engineering Plant Analyzer (EPA) had been developed in 1984 at Brookhaven National Laboratory to simulate plant transients in boiling water reactors (BWR). Recently, the EPA with its High-Speed Interactive Plant Analyzer code for BWRs ( ppercase HIPA-BWR ) simulated for the first time oscillatory transients with large, non-linear power and flow amplitudes; transients which are centered around the March 9, 1988 instability at the LaSalle-2 BWR power plant.The EPA's capability to simulate oscillatory transients has been demonstrated first by comparing simulation results with LaSalle-2 plant data (Wulff et al., NUREG/CR-5816, BNL-NUREG-52312, Brookhaven National Laboratory, 1992). This paper presents an EPA assessment on the basis of the Peach Bottom 2 instability tests (Carmichael and Niemi, EPRI NP-564, Electric Power Research Institute, Palo Alto, CA, 1978). This assessment of the EPA appears to constitute the first validation of a time-domain reactor systems code on the basis of frequency-domain criteria, namely power spectral density, gain and phase shift of the pressure-to-power transfer function.The reactor system pressure was disturbed in the Peach Bottom 2 power plant tests, and in their EPA simulation, by a pseudo-random, binary sequence signal. The data comparison revealed that the EPA predicted for Peach Bottom tests PT1, PT2, and PT4 the gain of the power-to-pressure transfer function with the biases and standard deviations of (-10±28)%, (-1±40)% and (+28±52)%, respectively. The respective frequencies at the peak gains were predicted with the errors of +6%, +3%, and -28%. The differences between the predicted and the measured phase shift increased with increasing frequency, but stayed within the margin of experimental uncertainty. ((orig.))

  1. IAEA provisional code of practice on management of radioactive waste from nuclear power plants

    International Nuclear Information System (INIS)

    1982-10-01

    This Code of Practice defines the minimum requirements for operations and design of structures, systems and components important for management of wastes from thermal nuclear power plants. It emphasizes what safety requirements shall be met rather than specifies how these requirements can be met; the latter aspect is covered in Safety Guides. The Code defines the need for a Government to assume responsibility for regulating waste management practices in conjunction with the regulation of a nuclear power plant. The Code does not prejudge the organization of the regulatory authority, which may differ from one Member State to another, and may involve more than one body. Similarly, the Code does not deal specifically with the functions of a regulatory authority responsible for such matters, although it may be of value to Member States in providing a basis for consideration of such functions. The Code deals with the entire management system for all wastes from nuclear power plants embodying thermal reactors including PWR, BWR, HWR and HTGR technologies. Topics included are: design, normal and abnormal operation, and regulation of management systems for gaseous, liquid and solid wastes, including decommissioning wastes. The Code includes measures to be taken with regard to the wastes arising from spent fuel management at nuclear power plants. However, the options for further management of spent fuel are only outlined since it is the subject of decisions by individual Member States. The Code does not require that an option(s) be decided upon prior to construction or operation of a nuclear power plant

  2. Flux and power distributions in BWR multi-bundle fuel arrays

    International Nuclear Information System (INIS)

    Cheng, H.S.

    1976-02-01

    Multi-bundle calculations have been performed in order to shed some light on an abnormal TIP trace recently discovered in a BWR/3. Transport theory was employed to perform the calculations with ENDF/B-IV data. The results indicate that a strong variation of the TIP reading does exist along the narrow water gap of a BWR due to the steep gradient of the thermal neutron flux; the maxima occurring at the intersections of the water gaps and the minima in between. Using this characteristic behavior of the TIP reading, together with the observed normal TIP trace, the abnormal behavior of the affected TIP trace exhibiting three peaks along the channel was roughly simulated. The calculations confirmed that the observed TIP trace anomaly was caused by the severe bending of the affected instrument tube as was actually discovered. The effect of hot water intrusion into the TIP guide tube, as well as that of loading the new 8 x 8 reload bundles, was also evaluated

  3. Development of membrane moisture separator for BWR off-gas system

    International Nuclear Information System (INIS)

    Ogata, H.; Kawamura, S.; Kumasaka, M.; Nishikubo, M.

    2001-01-01

    In BWR plant off-gas treatment systems, dehumidifiers are used to maintain noble gas adsorption efficiency in the first half of the charcoal hold-up units. From the perspective of simplifying and reducing the cost of such a dehumidification system, Japanese BWR utilities and plant fabricators have been developing a dehumidification system employing moisture separation membrane of the type already proven in fields such as medical instrumentation and precision measuring apparatus. The first part of this development involved laboratory testing to simulate the conditions found in an actual off-gas system, the results of which demonstrated satisfactory results in terms of moisture separation capability and membrane durability, and suggested favorable prospects for application in actual off-gas systems. Further, in-plant testing to verify moisture separation capability and membrane durability in the presence of actual gases is currently underway, with results so far suggesting that the system is capable of obtaining good moisture separation capability. (author)

  4. Timing criteria for supplemental BWR emergency response equipment

    International Nuclear Information System (INIS)

    Bickel, John H.

    2015-01-01

    The Great Tohuku Earthquake and subsequent Tsunami represented a double failure event which destroyed offsite power connections to Fukushima-Daiichi site and then destroyed on-site electrical systems needed to run decay heat removal systems. The accident could have been mitigated had there been supplemental portable battery chargers, supplemental pumps, and in-place piping connections to provide alternate decay heat removal. In response to this event in the USA, two national response centers, one in Memphis, Tennessee, and another in Phoenix, Arizona, will begin operation. They will be able to dispatch supplemental emergency response equipment to any nuclear plant in the U.S. within 24 hours. In order to define requirements for supplemental nuclear power plant emergency response equipment maintained onsite vs. in a regional support center it is necessary to confirm: (a) the earliest time such equipment might be needed depending on the specific scenario, (b) the nominal time to move the equipment from a storage location either on-site or within the region of a nuclear power plant, and (c) the time required to connect in the supplemental equipment to use it. This paper describes an evaluation process for a BWR-4 with a Mark I Containment starting with: (a) severe accident simulation to define best estimate times available for recovery based on the specific scenario, (b) identify the key supplemental response equipment needed at specific times to accomplish recovery of key safety functions, and (c) evaluate what types of equipment should be warehoused on-site vs. in regional response centers. (authors)

  5. Compilation of data concerning know and suspected water hammer events in nuclear power plants, CY 1969

    International Nuclear Information System (INIS)

    Chapman, R.L.; Christensen, D.D.; Dafoe, R.E.; Hanner, O.M.; Wells, M.E.

    1981-05-01

    This report compiles data concerning known and suspected water hammer events reported by BWR and PWR power plants in the United States from January 1, 1969, to May 1, 1981. This information is summarized for each event and is tabulated for all events by plant, plant type, year of occurrence, type of water hammer, system affected, basis/cause for the event, and damage incurred. Information is also included from other events not specifically identified as water hammer related. These other events involved vibration and/or system components similar to those involved in the water hammer events. The other events are included to ensure completeness of the report, but are not used to point out particular facts or trends. This report does not evaluate findings abstracted from the data

  6. MITI project on advanced man-machine system for nuclear power plants

    International Nuclear Information System (INIS)

    Kato, Kanji; Watanabe, Takao; Hayakawa, Hiroyasu; Naito, Norio; Masui, Takao; Ogino, Takamichi.

    1988-01-01

    A computerized operator support system (COSS) against abnormal plant conditions was developed as a five-year project from 1980 to 1984, under the sponsorship of the Ministry of International Trade and Industry. The main purpose of the COSS development was to implement the lessons learned from the Three Mile Island accident. The main nuclear industries in Japan participated in the project. The design concept of the operator support functions and the method to implement it were established, and the prototype systems of the COSS for BWR and PWR plants were developed. After the completion of the COSS development, the above participant group once again joined for the work on an advanced man-machine system for nuclear power plants (MMS-NPP). This eight-year project aims at realizing an advanced operator support system by applying artificial intelligence, especially knowledge engineering, and sophisticated man-machine interface devices. Its main objectives are shown. This system configuration, operating method decision system, man-machine communication system, operation and maintenance support functions and so on are described. (Kako, I.)

  7. LOSP-initiated event tree analysis for BWR

    International Nuclear Information System (INIS)

    Watanabe, Norio; Kondo, Masaaki; Uno, Kiyotaka; Chigusa, Takeshi; Harami, Taikan

    1989-03-01

    As a preliminary study of 'Japanese Model Plant PSA', a LOSP (loss of off-site power)-initiated Event Tree Analysis for a Japanese typical BWR was carried out solely based on the open documents such as 'Safety Analysis Report'. The objectives of this analysis are as follows; - to delineate core-melt accident sequences initiated by LOSP, - to evaluate the importance of core-melt accident sequences in terms of occurrence frequency, and - to develop a foundation of plant information and analytical procedures for efficiently performing further 'Japanese Model Plant PSA'. This report describes the procedure and results of the LOSP-initiated Event Tree Analysis. In this analysis, two types of event trees, Functional Event Tree and Systemic Event Tree, were developed to delineate core-melt accident sequences and to quantify their frequencies. Front-line System Event Tree was prepared as well to provide core-melt sequence delineation for accident progression analysis of Level 2 PSA which will be followed in a future. Applying U.S. operational experience data such as component failure rates and a LOSP frequency, we obtained the following results; - The total frequency of core-melt accident sequences initiated by LOSP is estimated at 5 x 10 -4 per reactor-year. - The dominant sequences are 'Loss of Decay Heat Removal' and 'Loss of Emergency Electric Power Supply', which account for more than 90% of the total core-melt frequency. In this analysis, a higher value of 0.13/R·Y was used for the LOSP frequency than experiences in Japan and any recovery action was not considered. In fact, however, there has been no experience of LOSP event in Japanese nuclear power plants so far and it is also expected that offsite power and/or PCS would be recovered before core melt. Considering Japanese operating experience and recovery factors will reduce the total core-melt frequency to less than 10 -6 per reactor-year. (J.P.N.)

  8. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  9. An A BWR demonstration simulator for training and developing technical staff

    International Nuclear Information System (INIS)

    Powers, J.; Yonezawa, H.; Aoyagi, Y.; Kataoka, K.

    2015-09-01

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  10. An A BWR demonstration simulator for training and developing technical staff

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  11. Prospects for nuclear power

    International Nuclear Information System (INIS)

    Kaplan, G.

    1983-01-01

    Describes how the nuclear power industry is improving plant operation and safety procedures and is reducing maintenance costs as it hopes for a brighter political climate. Points out that most of the efforts focus on key areas, such as improvements in control rooms and operator training and studies of physical processes within the reactor and associated systems. Discusses the increasing complexity of nuclear plants, the use of computers to process data in BWR plants, the decommissioning of old plants, and plant safety research activities worldwide. Offers an annotated bibliography

  12. Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

    International Nuclear Information System (INIS)

    Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

    1992-01-01

    The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

  13. Direct torus venting analysis for Chinshan BWR-4 plant with MARK-I containment

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2017-03-15

    Highlights: • Study the effectiveness of Direct Torus Venting System (DTVS) during extended SBO of 24 h for Chinshan MARK-I plant. • Containment response is analyzed by GOTHIC based on boundary conditions from RETRAN calculation. • Analyses are performed with and without DTVS, respectively. • Suppression pool is sub-divided and thermal stratification is observed. - Abstract: The Chinshan plant, owned by Taiwan Power Company, has twin units of BWR-4 reactor and MARK-I containment. Both units have been operating at rated core thermal power of 1840 MWt. The existing Direct Torus Venting System (DTVS) is the main system used for venting the containment during the extended station blackout event. The purpose of this paper is to study the effects of the DTVS venting on the response of the containment pressure and temperature. The reactor is depressurized by manually opening the safety relief valves (SRVs) during the SBO, which causes the mass and energy to be discharged into and heat up the suppression pool. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The DTVS model is established in the GOTHIC model based on the venting size, venting piping loss, venting initiation time, and venting source. The lumped volume model, 1-D coarse-mesh model, and 3-D coarse-mesh model are considered in the torus volume. The calculation is first done without DTVS venting to establish a reference basis. Then a case with DTVS available is performed. Comparison of the two cases shows that the existing DTVS design is effective in mitigating the severity of the containment pressure and temperature transients. The results also show that the 1-D coarse-mesh model may not be appropriate since a

  14. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  15. Experimental study on RCCV of ABWR plant-Part 1

    International Nuclear Information System (INIS)

    Saita, H.; Kikuchi, R.; Muramatsu, Y.; Hiramoto, M.; Oyamada, O.; Furukawa, H.; Sasagawa, K.; Ohmori, N.; Suzuki, S.; Suoita, M.; Kobayashi, I.; Yamaguchi, I.

    1989-01-01

    Steel containment vessels have generally been used for BWR (Boiling Water Reactor)-type nuclear power plants in Japan. However, in accordance with the development of the advanced-type BWR (ABWR), the introduction of a reinforced concrete containment vessel (RCCV) has been proposed because it enables greater freedom in flexibility of configuration. The study presented here accumulates basic data on RCCV's and verifies the integrity and ultimate strength of a trial-designed RCCV and the validity of its design method

  16. BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

    International Nuclear Information System (INIS)

    Sicard, D.; Verdier, A.; Monsigny, P.A.

    2004-01-01

    The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators

  17. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

    2014-05-15

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

  18. Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

    International Nuclear Information System (INIS)

    Park, Soo Young; Ahn, Kwang Il

    2012-01-01

    Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

  19. The use of the partial coherence function technique for the investigation of BWR noise dynamics

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1983-01-01

    The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

  20. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

    Greene, S.R.

    1986-01-01

    The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

  1. Seismic capacities of existing nuclear power plant structures

    International Nuclear Information System (INIS)

    Wesley, D.A.; Hashimoto, P.S.; Narver, R.B.

    1983-01-01

    The paper presents a discussion of the more important conservatisms and some of the results obtained when this methodology has been applied to various nuclear plants. Results are shown for both BWR and PWR plants, on both rock and soil sites, and for plants and soil sites, and for plants that were designed in the late 1960s to plants that have yet to load fuel. Safe shutdown earthquake design levels of 0.1 g to 0.25 g were used for these plants. Overall median structural factors of safety for the lowest significant seismic failure capacity at each plant ranged from 3.5 to 8.5. The lowest containment-related failure capacity at each plant ranged from 4.6 to 31. The types of failure corresponding to each safety factor are also tabulated. (orig./HP)

  2. The costs of nuclear power in the Netherlands

    International Nuclear Information System (INIS)

    1978-01-01

    A study on the costs of nuclear power generation in the Netherlands is presented. Light water cooled reactors are chosen as nuclear power plants and no difference is made in calculating the costs between a PWR type reactor and a BWR type reactor. The power plants have an output of 1000 MWe. From each part of the whole fuel cycle the costs are determined, taking into account interest, investments, time of construction, labor costs, insurances etc. Also are determined from each part of the fuel cycle the energy costs; the costs per kWh. Finally a comparison is made in costs between a 1000 MWe power plant and a 600 MWe power plant

  3. Integrated Level 3 risk assessment for the LaSalle Unit 2 nuclear power plant

    International Nuclear Information System (INIS)

    Payne, A.C. Jr.; Brown, T.D.; Miller, L.A.

    1991-01-01

    An integrated Level 3 probabilistic risk assessment (PRA) was performed on the LaSalle County Station nuclear power plant using state-of-the-art PRA analysis techniques. The objective of this study was to provide an estimate of the risk to the offsite population during full power operation of the plant and to include a characterization of the uncertainties in the calculated risk values. Uncertainties were included in the accident frequency analysis, accident progression analysis, and the source term analysis. Only weather uncertainties were included in the consequence analysis. In this paper selected results from the accident frequency, accident progression, source term, consequence, and integrated risk analyses are discussed and the methods used to perform a fully integrated Level 3 PRA are examined. LaSalle County Station is a two-unit nuclear power plant located 55 miles southwest of Chicago, Illinois. Each unit utilizes a Mark 2 containment to house a General Electric 3323 MWt BWR-5 reactor. This PRA, which was performed on Unit 2, included internal as well as external events. External events that were propagated through the risk analysis included earthquakes, fires, and floods. The internal event accident scenarios included transients, transient-induced LOCAs (inadvertently stuck open relief valves), anticipated transients without scram, and loss of coolant accidents

  4. Operating experience from Swedish nuclear power plants, 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The total generation of electricity from Swedish nuclear power plants was 70.1 TWh during 1999, which is slightly more than the mean value for the last five years. The total electricity consumption decreased by one percent, compared with 1998, to a total of 142.3 TWh, due to an unusually warm summer and autumn. The abundant supply of hydroelectric power resulted in comparatively extensive load-following operation by the nuclear plants during the year. Production losses due to low demand totalled 3.0 TWh. The closure of Barsebaeck 1 will result in a capacity reduction exceeding 4 TWh per year. The hydroelectric power production was 70 TWh, which was 6 TWh more than during a normal year, i.e. a year with average rainfall. The remaining production sources, mainly from solid fuel plants combined with district heating contributed 9 TWh. Electricity generation by means of wind power is still increasing. There are now about 470 wind power stations, which produced 0.3 TWh during the year. The total electricity generation totalled 149.8 TWh, a three percent decrease compared with 1998. The preliminary figures for export were 15.9 TWh and for import 8.4 TWh. The figures above are calculated from the preliminary production result. A comprehensive report on electric power supply and consumption in Sweden is provided in the 1999 Annual Report from the Swedish Power Association. The unit capability factor for the PWRs at Ringhals averaged 91%, while the BWRs averaged 82% mainly due to the extended outages. The BWR reactors at Forsmark averaged as much as 93%. Forsmark 1 experienced the shortest refuelling outage ever in Sweden, only 9 days and 20 hours. In May, Oskarshamn 2 passed a historical milestone - the unit produced 100 TWh since connection to the grid in 1974. The final production day for Barsebaeck 1, which had been in commercial operation since 1975, was on November 30 when a decision by the Swedish Government revoked the operating licence. Three safety-related events

  5. Analysis of BWR out-of-phase instabilities in the frequency domain

    International Nuclear Information System (INIS)

    Farawila, Y.M.; Pruitt, D.W.; Kreuter, D.

    1992-01-01

    During startup or because of an inadvertent recirculation pump trip, a boiling water reactor (BWR) may operate at relatively low flow and high power conditions. At these conditions, a BWR is susceptible to coupled flow and power oscillations that could result in undesirable reactor scram unless appropriate countermeasures are taken. This contribution to analytical methods has been developed to address in part a general industrywide and regulatory concern about BWR stability initiated by the LaSalle 2 instability event in March 1988. This work is designed to extend the capability of the one-dimensional parallel channel frequency domain code STAIF to predict the regional oscillation decay ratio. The basic theory follows that developed by March-Leuba and Blakeman, where the oscillation mechanism is identified as the excitation of a subcritical neutronic mode with a constant core pressure drop boundary condition. The improvements to the basic theory include applying the theory to one-dimensional neutronics instead of point kinetics and taking account of the actual three-dimensional harmonic flux distribution

  6. Noncondensable gas accumulation phenomena in nuclear power plant piping

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Aoki, Kazuyoshi; Sato, Teruaki; Shida, Akira; Ichikawa, Nagayoshi; Nishikawa, Akira; Inagaki, Tetsuhiko

    2011-01-01

    In the case of the boiling water reactor, hydrogen and oxygen slightly exist in the main steam, because these noncondensable gases are generated by the radiolytic decomposition of the reactor water. BWR plants have taken measures to prevent noncondensable gas accumulation. However, in 2001, the detonation of noncondensable gases occurred at Hamaoka-1 and Brunsbuttel, resulting in ruptured piping. The accumulation phenomena of noncondensable gases in BWR closed piping must be investigated and understood in order to prevent similar events from occurring in the future. Therefore, an experimental study on noncondensable gas accumulation was carried out. The piping geometries for testing were classified and modeled after the piping of actual BWR plants. The test results showed that 1) noncondensable gases accumulate in vertical piping, 2) it is hard for noncondensable gases to accumulate in horizontal piping, and 3) noncondensable gases accumulate under low-pressure conditions. A simple accumulation analysis method was proposed. To evaluate noncondensable gas accumulation phenomena, the three component gases were treated as a mixture. It was assumed that the condensation amount of the vapor is small, because the piping is certainly wrapped with heat insulation material. Moreover, local thermal equilibrium was assumed. This analysis method was verified using the noncondensable gas accumulation test data on branch piping with a closed top. Moreover, an experimental study on drain trap piping was carried out. The test results showed that the noncondensable gases dissolved in the drain water were discharged from the drain trap, and Henry's law could be applied to evaluate the amount of dissolved noncondensable gases in the drain water. (author)

  7. Quality control of stainless steel pipings for nuclear power generation

    International Nuclear Information System (INIS)

    Miki, Minoru; Kitamura, Ichiro; Ito, Hisao; Sasaki, Ryoichi

    1979-01-01

    The proportion of nuclear power in total power generation is increasing recently in order to avoid the concentrated dependence on petroleum resources, consequently the reliability of operation of nuclear power plants has become important. In order to improve the reliability of plants, the reliability of each machine or equipment must be improved, and for the purpose, the quality control at the time of manufacture is the important factor. The piping systems for BWRs are mostly made of carbon steel, and stainless steel pipings are used for the recirculation system cooling reactors and instrumentation system. Recently, grain boundary type stress corrosion cracking has occurred in the heat-affected zones of welded stainless steel pipings in some BWR plants. In this paper, the quality control of stainless steel pipings is described from the standpoint of preventing stress corrosion cracking in BWR plants. The pipings for nuclear power plants must have sufficient toughness so that the sudden rupture never occurs, and also sufficient corrosion resistance so that corrosion products do not raise the radioactivity level in reactors. The stress corrosion cracking occurred in SUS 304 pipings, the factors affecting the quality of stainless steel pipings, the working method which improves the corrosion resistance and welding control are explained. (Kako, I.)

  8. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  9. Water quality monitoring device for nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsushi.

    1995-01-01

    The device of the present invention measures quality of feedwater after heated in a regenerative heat exchanger device of a coolant cleanup system in a BWR type reactor, to detect ions generated from organic materials decomposed at high temperature and specify the position where impurities are formed. Namely, in a power plant having a reactor coolant cleanup pipeline connected to a feedwater pipeline, a water quality measuring portion is disposed to the feedwater system at the downstream of the junction to the feedwater system pipeline. A water quality sample is taken to measure the water quality in a state where the feedwater heated by a feedwater heater and flowing to the reactor, and the cleanup coolants heated by the regenerative heat exchanger are mixed. Thus, the impurities formed at the down stream of the feedwater system pipeline, as well as the water quality including impurities decomposed in a high temperature state can be measured. (I.S.)

  10. Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

    2009-10-15

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  11. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models; SUN-RAH: simulador universitario de nucleoelectrica BWR basado en modelos de orden reducido

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2003-07-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  12. 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

  13. Reactor power control device

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. Leibstadt: a 950-MW(e) BWR/6 Mark-III in commercial operation

    International Nuclear Information System (INIS)

    Fischer, P.U.

    1985-01-01

    It may be somewhat premature to report on a plant that started up in 1984 as the first of General Electric's (GE's) BWR/6 Mark-III plants in the Western Hemisphere and commenced commercial operation on December 15, 1984. The theme of the session certainly applies to the overall Swiss nuclear program and the search for excellence has been our ambition out of economic and energy supply necessities. Leibstadt came on line just in time to cover the needs of the Swiss consumers during the winter of 84/85. It has provided reliable service from the outset and operated during the extreme European cold wave in January 1985 without interruption. In 1985 the plant is expected to cover approx.15% of the electricity needs of Switzerland. The encouraging start of commercial operation gives hope that with time Leibstadt will be able to approach the capacity factors of the other four Swiss nuclear power stations, which in 1984 were between 88.4 and 90.3%

  15. Exxon Nuclear WREM-based NJP-BWR ECCS evaluation model and example application to the Oyster Creek Plant

    International Nuclear Information System (INIS)

    Krysinski, T.L.; Bjornard, T.A.; Steves, L.H.

    1975-01-01

    A proposed integrated ECCS model for non-jet pump boiling water reactors is presented, using the RELAP4-EM/BLOWDOWN and RELAP4-EM/SMALL BREAK portions of the Exxon Nuclear WREM-based Generic PWR Evaluation Model coupled with the ENC NJP-BWR Fuel Heatup Model. The results of the application of the proposed model to Oyster Creek are summarized. The results of the break size sensitivity study using the proposed model for the Oyster Creek Plant are presented. The application of the above results yielded the MAPLHGR curves. Included are a description of the proposed non-jet pump boiling water reaction evaluation model, justification of its conformance with TOCFR50, Appendix K, the adopted Oyster Creek plant model, and results of the analysis and sensitivity studies. (auth)

  16. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  17. Nuclear power plants

    International Nuclear Information System (INIS)

    1985-01-01

    Data concerning the existing nuclear power plants in the world are presented. The data was retrieved from the SIEN (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: nuclear plants, its status and type; installed nuclear power plants by country; nuclear power plants under construction by country; planned nuclear power plants by country; cancelled nuclear power plants by country; shut-down nuclear power plants by country. (E.G.) [pt

  18. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  19. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  20. BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

    International Nuclear Information System (INIS)

    Wattez, L.; Marguerat, Y.; Hoesli, C.

    2006-01-01

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN TM 24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN TM 9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN TM 9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN TM 24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN TM 9/4 round trips are performed, and one TN TM 24BH is loaded. 5 additional TN TM 24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN TM 24BH high capacity dual purpose cask and the TN TM 9/4 transport cask characteristics and capabilities will then be detailed. (authors)

  1. Construction works of large scale impervious wall in construction of No.2 plant in Onagawa Nuclear Power Station, Tohoku Electric Power Co., Inc

    International Nuclear Information System (INIS)

    Ueda, Kozaburo; Sugeno, Yoshisada; Takahashi, Hitoshi

    1991-01-01

    The main buildings for No. 2 plant in Onagawa Nuclear Power Station are constructed on the bedrocks about 14 m below the sea surface. Therefore, for the purpose of executing the works by shutting seawater off and dry work, the large scale impervious wall of about 500 m extension was installed underground. The feature of this impervious wall is the depth of embedment of about 3 m into the hard bedrocks having the uniaxial compressive strength of 2000 kg/cm 2 at maximum, carried out with the newly developed hard rock excavator. The outline of these construction works is reported. No. 2 plant in Onagawa Nuclear Power Station is the BWR plant of 825 MWe output. The construction works of the power station were began in August, 1989, and the rate of progress in civil engineering works as of the end of September, 1990 was 21.3%. The planning of the impervious wall, the geological features at the site, the method of shutting seawater off, the selection of wall materials, the design of the wall body, the investigation of the quantity of spring water, the execution of the construction and execution management, and the confirmation of the effect of the wall are reported. (K.I.)

  2. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  3. BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

    International Nuclear Information System (INIS)

    Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

    1994-04-01

    A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137 Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

  4. Searching the beginning of BWR power instability events with the Hilbert Huang transform

    International Nuclear Information System (INIS)

    Blázquez, Juan; Montalvo, Cristina; García-Berrocal, Agustín; Balbás, Miguel

    2013-01-01

    Highlights: ► The report of the instability is enriched by including its beginning and its end. ► The Hilbert Huang transform (HHT) is used for indentifying both. ► The first Intrinsic Mode Function (IMF) detects both. ► The methodology is applied to neutron detector signals from two plants. ► The Decay Ratio of IMF 1 is calculated. - Abstract: When a BWR instability takes place, the Regulator usually demands a report which must include many aspects such as the initial time of the instability and also the measurements adopted by the operator at that time. This initial time normally is difficult to know from the available data. In this work, a methodology is proposed to determine accurately when the instability began based on the Hilbert–Huang transform. The Empirical Mode Decomposition is applied to neutron detector signals coming from two plants which have recorded them during real instability events. The first intrinsic mode function shows sharply the beginning and the end of the incident. Besides, through the instantaneous amplitude and frequency of the first mode a kind of Decay Ratio can be assigned allowing us to obtain a sharper description of the instability

  5. OECD/NRC BWR Turbine Trip Benchmark: Simulation by POLCA-T Code

    International Nuclear Information System (INIS)

    Panayotov, Dobromir

    2004-01-01

    Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and three-dimensional (3-D) neutron kinetics core models. Participation in the OECD/NRC BWR Turbine Trip (TT) Benchmark is a part of our efforts toward the code's validation. The paper describes the objectives for TT analyses and gives a brief overview of the developed plant system input deck and 3-D core model.The results of exercise 1, system model without netronics, are presented. Sensitivity studies performed cover the maximal time step, turbine stop valve position and mass flow, feedwater temperature, and steam bypass mass flow. Results of exercise 2, 3-D core neutronic and thermal-hydraulic model with boundary conditions, are also presented. Sensitivity studies include the core inlet temperature, cladding properties, and direct heating to core coolant and bypass.The entire plant model was validated in the framework of the benchmark's phase 3. Sensitivity studies include the effect of SCRAM initialization and carry-under. The results obtained - transient fission power and its initial axial distribution and steam dome, core exit, lower and upper plenum, main steam line, and turbine inlet pressures - showed good agreement with measured data. Thus, the POLCA-T code capabilities for correct simulation of pressurizing transients with very fast power were proved

  6. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  7. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    Lopez R, A.

    2004-01-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  8. LNG plant combined with power plant

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, I; Kikkawa, Y [Chiyoda Chemical Engineering and Construction Co. Ltd., Tokyo (Japan)

    1997-06-01

    The LNG plant consumers a lot of power of natural gas cooling and liquefaction. In some LNG plant location, a rapid growth of electric power demand is expected due to the modernization of area and/or the country. The electric power demand will have a peak in day time and low consumption in night time, while the power demand of the LNG plant is almost constant due to its nature. Combining the LNG plant with power plant will contribute an improvement the thermal efficiency of the power plant by keeping higher average load of the power plant, which will lead to a reduction of electrical power generation cost. The sweet fuel gas to the power plant can be extracted from the LNG plant, which will be favorable from view point of clean air of the area. (Author). 5 figs.

  9. LNG plant combined with power plant

    International Nuclear Information System (INIS)

    Aoki, I.; Kikkawa, Y.

    1997-01-01

    The LNG plant consumers a lot of power of natural gas cooling and liquefaction. In some LNG plant location, a rapid growth of electric power demand is expected due to the modernization of area and/or the country. The electric power demand will have a peak in day time and low consumption in night time, while the power demand of the LNG plant is almost constant due to its nature. Combining the LNG plant with power plant will contribute an improvement the thermal efficiency of the power plant by keeping higher average load of the power plant, which will lead to a reduction of electrical power generation cost. The sweet fuel gas to the power plant can be extracted from the LNG plant, which will be favorable from view point of clean air of the area. (Author). 5 figs

  10. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  11. Research on U.S. nuclear power plant major equipment aging

    International Nuclear Information System (INIS)

    Nakos, J.T.; Rosinski, S.T.

    1994-01-01

    The U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), in cooperation with nuclear power plant utilities and the Nuclear Energy Institute, have prepared equipment aging evaluations of nuclear power plant equipment for life extension considerations. Specifically, these evaluations focused on equipment considered important for plant license renewal (U.S. Code of Federal Regulations 10CFR54). open-quotes Industry Reportsclose quotes (IRs), jointly funded by DOE and EPRI, evaluated the aging of major systems, structures, and components (e.g., reactor pressure vessels, Class I structures, PWR and BWR containments, etc.) and contain a mixture of technical and licensing information. open-quotes Aging Management Guidelinesclose quotes (AMGs), funded by DOE, evaluate aging for commodity types of equipment (e.g., pumps, electrical switchgear, heat exchangers, etc.) and concentrate on technical issues only. AMGs are intended for systems engineers and plant maintenance staff. A significant number of technical issues were resolved during IR interactions with the U.S. Nuclear Regulatory Commission (NRC). However, certain technical issues have not been resolved and are considered open-quotes openclose quotes. Examples include certain issues related to fatigue, neutron irradiation embrittlement, intergranular stress corrosion cracking (IGSCC) and electrical cable equipment qualification. Direct NRC interaction did not take place during preparation of individual AMGs due to their purely technical nature. The eventual use of AMGs in a future license renewal application will likely require NRC interaction at that time. With a few noted exceptions, the AMG process indicated that current aging management practices of U.S. utilities were effective in preventing age-related degradation. This paper briefly describes the IR and AMG processes and summarizes the unresolved technical issues identified through preparation of the documents

  12. Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Williams, T. [EGL Laufenburg (Switzerland); Helmersson, S. [Westinghouse Atom (Sweden)

    2001-03-01

    The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

  13. Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

    International Nuclear Information System (INIS)

    Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R.; Williams, T.; Helmersson, S.

    2001-01-01

    The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

  14. Qualification, certification and training systems of a Japanese nuclear power plant supplier

    International Nuclear Information System (INIS)

    Yoshijima, S.; Tomita, J.; Yoneda, Y.

    1982-01-01

    Training and indoctrination of personnel are very important for performing quality assurance activities in nuclear power plants. The paper presents policies, procedures and practices with regard to a training system for site construction workers, a training system for plant operation personnel at a special facility, and services and activities for operating plants. The training system for site workers and technicians offers basic training for all workers, and special training and qualification for field welders, cable-termination workers and non-destructive examination personnel. In addition to the qualification system for field welders applied by the regulatory body, a privat-company qualification system exists. Also, a training centre for BWR operators has been established. This facility has a simulator duplicating the main control room of an actual plant and a computer-assisted instruction system. Standard training courses, short-term basic courses, re-training courses, group training courses and special training courses are held at the training centre. Finally, the services and activities performed by Toshiba Corporation for operating plants are described. These activities of the plant supplier aim at keeping up and further enhancing the safety and reliability of operating plants and mainly consist of: (1) collection and evaluation of plant operation data, with re-evaluation and improvement of systems and components; (2) development of new maintenance techniques; (3) development of measures for reducing annual outage periods; and (4) plant emergency preparedness. (author)

  15. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  16. Report on the BWR owners group radiation protection/ALARA Committee

    International Nuclear Information System (INIS)

    Aldrich, L.R.

    1995-01-01

    Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming open-quotes World Classclose quotes performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance

  17. Report on the BWR owners group radiation protection/ALARA Committee

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, L.R. [Commonwealth Edison Co., Downers Grove, IL (United States)

    1995-03-01

    Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.

  18. Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Rodenas, J.; Abarca, A.; Gallardo, S.

    2011-01-01

    BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool.

  19. Current situation of spent fuel management in the Laguna Verde Nuclear Power Plant, Veracruz, Mexico

    International Nuclear Information System (INIS)

    Moreno, C.V.

    1994-01-01

    The Comision Federal de Electricidad (CFE), owner and operator of the Laguna Verde nuclear power plant (2 x 654 MWe BWR), has twice decided to increase the storage capacity of the spent fuel pools of the reactors. The Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS), the national nuclear regulatory authority, approved the increase by a factor of 2.66 in the storage capacity proposal by CFE in 1989. Each reactor spent fuel pool can now hold 614 t HM. The reracking was done at a cost of about US $13 per kg U, which will add only 0.042 mills per kWh to the fuel cycle cost. (author)

  20. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  1. Optimization of fuel reloads for a BWR using the ant colony system

    International Nuclear Information System (INIS)

    Esquivel E, J.; Ortiz S, J. J.

    2009-10-01

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  2. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  3. 15. Internal symposium on recent progress of nondestructive inspection and monitoring technologies for nuclear power plants

    International Nuclear Information System (INIS)

    1994-01-01

    At the symposium, lectures were given on the recent development of the nondestructive inspection technology for nuclear power plants, the trend regarding the nondestructive inspection in foreign countries (Japan-Germany atomic energy seminar), the present state and subjects of the monitoring technology in BWR plants, the present state and subjects of the monitoring technology in PWR plants, and the present state and the subjects for hereafter of the defect evaluation method in the equipment of light water reactors. The data on the ultrasonic flaw detection in aluminum alloy welded joints were obtained. The German inspection technology is similar to that in Japan and other countries. The research on the plant synthetic monitoring and diagnosis system is reported. The monitoring systems for abnormal state in operation, troubles and the secular change of equipment are reported. The evaluation of the flaws in nuclear piping is reported. The summaries of the lectures are collected in this book. (K.I.)

  4. ASME section XI: rules for inservice inspection of nuclear power plants -an introspection

    Energy Technology Data Exchange (ETDEWEB)

    John, P K; Anto, Y; Mungikar, C P; Wagh, P M [Nuclear Power Corporation of India Ltd., Tarapur (India). Tarapur Atomic Power Station

    1994-12-31

    Section XI of the ASME BPV code is addressed to the examination, test and inspection requirements of the components of nuclear power plants (NPPs). Since its inception in 1970, this code section has undergone vast changes -probably the most among other ASME BPV code sections. Section XI is full fledged and lays down requirements with regard to all preservice inspections, inservice inspection, repair and replacement of components, tests of system etc. Tarapur Atomic Power Station (TAPS) has the distinction of being one of the earliest BWR type NPPs in the world that has an inservice inspection programme organised in line with the ASME section XI requirements. This paper summarises the experiences gained from time to time using this code section and a few suggestions to prospective users. An effort is also made to explain the philosophy of inservice inspection from ASME section XI point of view. 3 refs.

  5. ASME section XI: rules for inservice inspection of nuclear power plants -an introspection

    International Nuclear Information System (INIS)

    John, P.K.; Anto, Y.; Mungikar, C.P.; Wagh, P.M.

    1994-01-01

    Section XI of the ASME BPV code is addressed to the examination, test and inspection requirements of the components of nuclear power plants (NPPs). Since its inception in 1970, this code section has undergone vast changes -probably the most among other ASME BPV code sections. Section XI is full fledged and lays down requirements with regard to all preservice inspections, inservice inspection, repair and replacement of components, tests of system etc. Tarapur Atomic Power Station (TAPS) has the distinction of being one of the earliest BWR type NPPs in the world that has an inservice inspection programme organised in line with the ASME section XI requirements. This paper summarises the experiences gained from time to time using this code section and a few suggestions to prospective users. An effort is also made to explain the philosophy of inservice inspection from ASME section XI point of view. 3 refs

  6. 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)

  7. Plant analyzer for high-speed interactive simulation of BWR plant transients

    International Nuclear Information System (INIS)

    Cheng, H.S.; Lekach, S.V.; Mallen, A.N.; Wulff, W.; Cerbone, R.J.

    1984-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology was utilized to develop a plant analyzer which affords realistic predictions of plant transients and severe off-normal events in LWR power plants through on-line simulations at speeds up to 10 times faster than actual process speeds. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the entire balance of the plant. Reactor core models include point kinetics with reactivity feedback due to void fraction, fuel temperature, coolant temperature, and boron concentration as well as a conduction model for predicting fuel and clad temperatures. Control systems and trip logic for plant protection systems are also simulated. The AD10 of Applied Dynamics International, a special-purpose peripheral processor, is used as the principal hardware of the plant analyzer

  8. Investigation of power oscillation mechanisms based on noise analysis at Forsmark-1 BWR

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1996-01-01

    Noise analysis has been performed for stability test data collected during reactor start-up in January 1989 at the boiling water reactor (BWR) Forsmark unit 1. A unique instrumentation to measure local coolant flow in this reactor allowed investigation of dynamic interactions between neutron flux and coolant flow noise signals at different radial positions in the core. The causal relationship for these signals was evaluated based on a method called signal transmission path (STP) analysis with the aim of identifying the principal mechanism of power oscillations in this reactor. The results of the present study indicated that large amplitude power oscillations were induced by two instability mechanisms concurrent in the core. The first is the global void reactivity feedback effect which played the most significant role to power oscillations at a resonant frequency of about 0.53 Hz. The second is the thermal-hydraulics coupling with neutron kinetics, inducing resonant oscillations at about 0.45 Hz. The latter was found to be active only in a certain core region. A peculiar phenomenon of amplitude modulations observed in some local power range monitor (LPRM) signals was also examined. It was interpreted to occur as the consequence of these two resonant power oscillations, the frequencies of which lie close to each other. The noise analysis technique applied in the present study is expected to be useful to get a deeper understanding of the power oscillation mechanism which is active in the reactor under evaluation. The technique may be applicable to BWRs with instruments to measure local channel flow together with in-core neutron detectors. (Author)

  9. An approach to review design bases and safety analysis of earlier generation atomic power plants; a case study of TAPS

    International Nuclear Information System (INIS)

    Malhotra, P.K.; Bajaj, S.S.

    2002-01-01

    The twin unit boiling water reactor (BWR) station at TAPS has completed 30 years of power operation and for further extending plant operating life, a fresh extensive exercise involving review of plant operating performance, aging management and review of design bases and safety analysis has been carried out. The review exercise resulted in assessment of acceptability of identified non-conformances and recommendation for compensatory measures in the form of design modification or plant operating procedures. The second part of the exercise is related to safety analysis, which is carried out in view of the plant modifications done and advances taken place in methodologies of analytical techniques. Chiefly, it involves LOCA analysis done for various break sizes at different locations and plant transient studies. It also includes the fatigue analysis of the reactor pressure vessel. The related review approach adopted is presented here

  10. Technical and economic proposal for the extension of the Laguna Verde Nuclear Power plant with an additional nuclear reactor

    International Nuclear Information System (INIS)

    Leal C, C.D.; Francois L, J.L.

    2006-01-01

    The increment of the human activities in the industrial environments and of generation of electric power, through it burns it of fossil fuels, has brought as consequence an increase in the atmospheric concentrations of the calls greenhouse effect gases and, these in turn, serious repercussions about the environment and the quality of the alive beings life. The recent concern for the environment has provoked that industrialized countries and not industrialized carry out international agreements to mitigate the emission from these gases to the atmosphere. Our country, like part of the international community, not is exempt of this problem for what is necessary that programs begin guided toward the preservation of the environment. As for the electric power generation, it is indispensable to diversify the sources of primary energy; first, to knock down the dependence of the hydrocarbons and, second, to reduce the emission of polluting gases to the atmosphere. In this item, the nucleo electric energy not only has proven to be safe and competitive technical and economically, able to generate big quantities of electric power with a high plant factor and a considerable cost, but rather also, it is one of the energy sources that less pollutants it emits to the atmosphere. The main object of this work is to carry out a technical and economic proposal of the extension of the Laguna Verde Nuclear power plant (CNLV) with a new nuclear reactor of type A BWR (Advanced Boiling Water Reactor), evolutionary design of the BWR technology to which belong the two reactors installed at the moment in the plant, with the purpose of increasing the installed capacity of generation of the CNLV and of the Federal Commission of Electricity (CFE) with foundation in the sustainable development and guaranteeing the protection of the environment by means of the exploitation of a clean and sure technology that counts at the moment with around 12,000 year-reactor of operational experience in more of

  11. An analysis of containment venting as a severe accident mitigation strategy for the BWR Mark II containment

    International Nuclear Information System (INIS)

    Kelly, D.L.; Galyean, W.J.

    1990-01-01

    An evaluation of a BWR/4 reactor with a Mark-II containment has identified the effects of containment venting on core damage frequency and containment failure mode, and has performed a limited evaluation of the effects on the off-site consequences. The analysis was founded upon an existing probabilistic risk assessment (PRA) with the addition of a proposed filtered containment venting system, based on the Swedish Filtra system installed at the Barseback nuclear power station in southern Sweden. Three different containment venting strategies were examined for their effects on plant risk. These are discussed

  12. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  13. Results of Cable Aging Management Tests for Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Garcia Hernandez, E.E.; Vazquez Cervantes, R.M.; Bonifacio M, J.; Garcia Garcia, J.

    2012-01-01

    Laguna Verde Nuclear Power Plant (LVNPP) located in Veracruz, Mexico is a BWR plant, two Units with 810 MWe each one, Unit 1 (1989) and Unit 2 (1990). The Equipment Qualification (EQ) Group at the Nuclear Research National Institute (ININ) has been working with the plant on tasks to develop the LVNPP cables Aging Management Program (AMP), as part of the technical basis to extend the operational life of the plant through license renewal up to 60 years. LVNPP cables are qualified for 40 years plus a LOCA DBA in accordance with 10.CFR 50.49 and the IEEE Std-323 and IEEE St. 383. The first studies for cables AMP have been performed with samples of safety related I and C cables taken from the LVNPP warehouse, similar brands and models as installed at the plant. ININ applied the condition monitoring techniques to these samples to identify predictive degradation and to establish the methodology for cables AMP, focused to the LVNPP license renewal. Cable tests program has been running at the EQ Lab in ININ, performing accelerated aging by steps up to 60 years and to 40 years plus a LOCA test. Determination for Activation Energy (Ea) and Oxidation Induction Time (OIT) methods were developed applying a DSC/TGA calorimeter. (author)

  14. Seismic proving tests on the reliability for large components and equipment of nuclear power plants

    International Nuclear Information System (INIS)

    Ohno, Tokue; Tanaka, Nagatoshi

    1988-01-01

    Since Japan has destructive earthquakes frequently, the structural reliability for large components and equipment of nuclear power plants are rigorously required. They are designed using sophisticated seismic analyses and have not yet encountered a destructive earthquake. When nuclear power plants are planned, it is very important that the general public understand the structural reliability during and after an earthquake. Seismic Proving Tests have been planned by Ministry of International Trade and Industry (Miti) to comply with public requirement in Japan. A large-scale high-performance vibration table was constructed at Tasted Engineering Laboratory of Nuclear Power Engineering Test Center (NU PEC), in order to prove the structural reliability by vibrating the test model (of full scale or close to the actual size) in the condition of a destructive earthquake. As for the test models, the following four items were selected out of large components and equipment important to the safety: Reactor Containment Vessel; Primary Coolant Loop or Primary Loop Recirculation System; Reactor Pressure Vessel; and Reactor Core Internals. Here is described a brief of the vibration table, the test method and the results of the tests on PWR Reactor Containment Vessel and BWR Primary Loop Recirculation System (author)

  15. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  16. Condensate treatment in BWR circuits by filter demineralizer units using powdered ion exchange resin at medium and high temperature

    International Nuclear Information System (INIS)

    De Martino, R.

    1983-01-01

    Considering the radiation build-up in some BWR reactors, we make a correlation between this phenomenon and the condensate purification system applied and the point of its utilization into the circuits. The application temperature of such a plant seems to have a very important role on the equilibria of metals contained in the reactor water and on the oxide composition. The efficiency of the condensate polishing system and the corrosion control are the most interesting objectives to achieve and to maintain, to control and regulate the physical and chemical process in the feedwater and in the reactor water. Up to date the technology owns major knowledge and a consistent know-how on using chemical products in order to increase the condensate polishing system efficiency. It is also considered a typical parallel case of a conventional power station and a secondary system of BWR units. (author)

  17. Development of neural network simulating power distribution of a BWR fuel bundle

    International Nuclear Information System (INIS)

    Tanabe, A.; Yamamoto, T.; Shinfuku, K.; Nakamae, T.

    1992-01-01

    A neural network model is developed to simulate the precise nuclear physics analysis program code for quick scoping survey calculations. The relation between enrichment and local power distribution of BWR fuel bundles was learned using two layers neural network (ENET). A new model is to introduce burnable neutron absorber (Gadolinia), added to several fuel rods to decrease initial reactivity of fresh bundle. The 2nd stages three layers neural network (GNET) is added on the 1st stage network ENET. GNET studies the local distribution difference caused by Gadolinia. Using this method, it becomes possible to survey of the gradients of sigmoid functions and back propagation constants with reasonable time. Using 99 learning patterns of zero burnup, good error convergence curve is obtained after many trials. This neural network model is able to simulate no learned cases fairly as well as the learned cases. Computer time of this neural network model is about 100 times faster than a precise analysis model. (author)

  18. AREVA 10x10 BWR fuel experience feedback and on going upgrading

    International Nuclear Information System (INIS)

    Lippert, Hans Joachim; Rentmeister, Thomas; Garner, Norman; Tandy, Jay; Mollard, Pierre

    2008-01-01

    Established with engineering and manufacturing operations in the US and Europe, AREVA NP has been and is supplying nuclear fuel assemblies and associated core components to boiling water reactors worldwide, representing today more than 63 000 fuel assemblies. The evolution of BWR fuel rod arrays from early 6x6 designs to the 10x10 designs first introduced in the mid 1990's yielded significant improvements in thermal mechanical operating limits, critical power level, cold shutdown margin, discharge burnup, as well as other key operational capabilities. Since first delivered in 1992, ATRIUM T M 1 0 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. This article presents in detail the operational experience consolidated by these more than 20 000 ATRIUM T M 1 0 BWR assemblies already supplied to utilities. Within the different 10x10 fuel assemblies available, the Fuel Assembly design is chosen and tailored to the operating strategies of each reactor. Among them, the latest versions of ATRIUM T M a re ATRIUM T M 1 0XP and ATRIUM T M 1 0XM fuel assemblies which have been delivered to several utilities worldwide. The article details key aspects of ATRIUM T M 1 0 fuel assemblies in terms of reliability and performance. Special attention is paid to key proven features, ULTRAFLOW T M s pacer grids, the use of part length fuel rods (PLFRs) and their geometrical optimization, water channel and load chain, upgraded features available for inclusion with most advanced designs. Regular upgrading of the product has been made possible thanks to a continuous improvement process with the aim of further upgrading BWR fuel assembly performance and reliability. Regarding thermal mechanical behavior of fuel rods, chromia (Cr2O3) doped fuel pellets, described in Reference 1, well illustrate this improvement strategy to reduce fission gas release, increase power thresholds for PCI

  19. Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

    International Nuclear Information System (INIS)

    Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

    2009-01-01

    The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

  20. 3D pin-by-pin power density profiles with high spatial resolution in the vicinity of a BWR control blade tip simulated with coupled neutronics/burn-up calculations

    International Nuclear Information System (INIS)

    Li, J.; Nünighoff, K.; Allelein, H.-J.

    2011-01-01

    Highlights: ► High spatial resolution neutronic and burn-up calculations of quarter BWR fuel element section. ► Coupled MCNP(X)–ORIGEN2.2 simulation using VESTA. ► Control blade history effect was taken into account. ► Determining local power excursion after instantaneous control rod movement. ► Correlation between control blade geometry and occurrence of local power excursions. - Abstract: Pellet cladding interaction (PCI) as well as pellet cladding mechanical interaction (PCMI) are well-known fuel failures in light water reactors, especially in boiling water reactors (BWR). Whereas the thermo-mechanical processes of PCI effects have been intensively investigated in the last decades, only rare information is available on the role of neutron physics. However, each power transient is primary due to neutron physics effects and thus knowledge of the neutron physical background is mandatory to better understand the occurrence of PCI effects in BWRs. This paper will focus on a study of local power excursions in a typical BWR fuel assembly during control rod movements. Burn-up and energy deposition were simulated with high spatial granularity, especially in the vicinity of the control blade tip. It could be shown, that the design of the control blade plays a dominant role for the occurrence of local power peaks while instantaneously moving down the control rod. The main result is, that the largest power peak occurs at the interface between steel handle and absorber rods. A full width half maximum (FWHM) of ±2.5 cm was observed. This means, the local power excursion due to neutron physics phenomena involve approximately five pellets. With the VESTA code coupled MCNP(X)/ORIGEN2.2 calculations were performed with more than 3400 burn-up zones in order to take history effects into account.