WorldWideScience

Sample records for nuclear plant evaluation

  1. Problems of nuclear power plant safety evaluation

    International Nuclear Information System (INIS)

    Suchomel, J.

    1977-01-01

    Nuclear power plant safety is discussed with regard to external effects on the containment and to the human factor. As for external effects, attention is focused on shock waves which may be due to explosions or accidents in flammable material transport and storage, to missiles, and to earthquake effects. The criteria for evaluating nuclear power plant safety in different countries are shown. Factors are discussed affecting the reliability of man with regard to his behaviour in a loss-of-coolant accident in the power plant. Different types of PWR containments and their functions are analyzed, mainly in case of accident. Views are discussed on the role of destructive accidents in the overall evaluation of fast reactor safety. Experiences are summed up gained with the operation of WWER reactors with respect to the environmental impact of the nuclear power plants. (Z.M.)

  2. Evaluation of nuclear power plant operator's ability

    International Nuclear Information System (INIS)

    Wei Li; He Xuhong; Zhao Bingquan

    2004-01-01

    Based on the quantitative research on nuclear power plant (NPP) operator's psychological characteristics and performance, the Borda's method of fuzzy mathematics combined with the character of operator's task is used to evaluate their abilities. The result provides the reference for operator's reliability research and psychological evaluation. (author)

  3. Seismic evaluation of existing nuclear power plants

    International Nuclear Information System (INIS)

    2003-01-01

    The IAEA nuclear safety standards publications address the site evaluation and the design of new nuclear power plants (NPPs), including seismic hazard assessment and safe seismic design, at the level of the Safety Requirements as well as at the level of dedicated Safety Guides. It rapidly became apparent that the existing nuclear safety standards documents were not adequate for handling specific issues in the seismic evaluation of existing NPPs, and that a dedicated document was necessary. This is the purpose of this Safety Report, which is written in the spirit of the nuclear safety standards and can be regarded as guidance for the interpretation of their intent. Worldwide experience shows that an assessment of the seismic capacity of an existing operating facility can be prompted for the following: (a) Evidence of a greater seismic hazard at the site than expected before, owing to new or additional data and/or to new methods; (b) Regulatory requirements, such as periodic safety reviews, to ensure that the plant has adequate margins for seismic loads; (c) Lack of anti-seismic design or poor anti-seismic design; (d) New technical finding such as vulnerability of some structures (masonry walls) or equipment (relays), other feedback and new experience from real earthquakes. Post-construction evaluation programmes evaluate the current capability of the plant to withstand the seismic concern and identify any necessary upgrades or changes in operating procedures. Seismic qualification is distinguished from seismic evaluation primarily in that seismic qualification is intended to be performed at the design stage of a plant, whereas seismic evaluation is intended to be applied after a plant has been constructed. Although some guidelines do exist for the evaluation of existing NPPs, these are not established at the level of a regulatory guide or its equivalent. Nevertheless, a number of existing NPPs throughout the world have been and are being subjected to review of their

  4. Research on psychological evaluation method for nuclear power plant operators

    International Nuclear Information System (INIS)

    Fang Xiang; He Xuhong; Zhao Bingquan

    2007-01-01

    The qualitative and quantitative psychology evaluation methods to the nuclear power plant operators were analyzed and discussed in the paper. The comparison analysis to the scope and result of application was carried out between method of outline figure fitted and method of fuzzy synthetic evaluation. The research results can be referenced to the evaluation of nuclear power plant operators. (authors)

  5. Economic evaluation of bids for nuclear power plants

    International Nuclear Information System (INIS)

    1976-01-01

    The purpose of the guidebook is to assist an organisation responsible for a nuclear power project in evaluating and establishing an economic order of merit among competing bids. An approximate overall time schedule for a first nuclear power plant project is provided. A schematic outline of technical bid evaluation is given. The basic procedure of economic bid evaluation is outlined, e.g. evaluation of the present worth of all cost items of plant capital investment, of the nuclear cycle, of O and M costs (operation and maintenance costs), and of economic corrections. All these cost items are evaluated for the economic life of the plant and corrected for escalation where applicable

  6. Actions concerning nuclear power plant life evaluation

    International Nuclear Information System (INIS)

    Chocron, M.; Fabbri, S.; Mizrahi, R.; Savino, E.J.; Versaci, R.A.

    1998-01-01

    One of the main activities to be undertaken by CNEA will be to provide technological assistance to NASA in problems concerning NPP operation. Works on life extensions of NPP are included in these activities. To fulfill these requirements the Atomic Energy National Commission (CNEA) has constituted a technical committee for Nuclear Power Plants Support (CAPCEN). CAPCEN should be the knowledge reservoir of those issues concerning the performance, safety and life extension of Nuclear Power Plants. One of CAPCEN's most important activities is to promote research work connected with such issues. The main technical areas are: Pressure Vessel and Piping, Heat Exchanges and Fuel Channels and Reactor Inner Components. Efforts are focused on the identification of the main components susceptible of ageing, the study of their ageing mechanisms, the follow-up of their behaviour during operation, and the measures taken to extend their life. (author)

  7. Evaluation of Visual and Landscape Impacts of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Algohary, S.A.

    2007-01-01

    The nuclear power plant is a huge structure, and in terms of both size and function may result in an unacceptable visual conflict in both local and wider environment. Also, it has major implications in terms of physical, social, economic, environmental and impact on people. The environmental impacts include the visual and landscape aspects of these plants. This paper outlines the main general ideas of the architecture aspects of nuclear power plants, nuclear reactors. Also, it discusses the site selection considerations: Finally, it introduces an approach for the evaluation of visual and landscape impacts of nuclear power plants

  8. Water hammer in USA nuclear power plants and it's evaluation

    International Nuclear Information System (INIS)

    Liu Shuqian.

    1987-01-01

    The results of evaluations about the water hammer events in USA nuclear power plants in recent years are summarily reported. The evaluations included underlying causes and frequency of water hammer events, damages incurred and systems affected. Through the evaluations about water hammer events and on the basis of past operation experiences in nuclear power plants, the design and operational modifications to prevent or mitigate water hammer events were presented. The NRC's current opinions relating to the water hammer problems are summarized, the importance of water hammer events for nuclear power plants construction in China is indicated

  9. Development of supplier evaluation model applying in nuclear power plants

    International Nuclear Information System (INIS)

    Wang Yonggang; Fang Chunfa

    2006-01-01

    It is essential for the safe and stable operations of Nuclear Power Plants that various resources in the supply chain are effectively managed. Supplier is a significant resource of nuclear entities serving as an extension of the operation process. Scientific and radiation evaluation of the performance of suppliers is of vital importance to an effective and high quality supply chain. This paper establishes an advance and practical supplier evaluation system that is applicable for the operational nuclear power plants, based on the analysis of the current operation status of Daya Bay Nuclear Power Station against its targeted objectives, the acquisition of relevant practices home and abroad and the benchmarking with advanced peers, in order to enhance the core competence of nuclear power plant. (authors)

  10. Method discussion of the performance evaluation on nuclear plant cable

    International Nuclear Information System (INIS)

    Lu Yongfang; Zhong Weixia; Sun Jiansheng; Liu Jingping

    2014-01-01

    A stock cable, which is same as the nuclear plant cable in service, was treated by thermal aging. After that, the mechanical property, the flame retardant property, the anti-oxidation were measured, and relationships between them due to the thermal aging were established. By those analysis, evaluating the in-service cable performance in nuclear plant and calculating its remaining life. Furthermore, the feasibility of this method was disscussed. (authors)

  11. Systematic evaluation of safety in old nuclear power plants

    International Nuclear Information System (INIS)

    Dredemis, G.; Fourest, B.

    1984-11-01

    A systematic evaluation program of the early nuclear power plant has been conducted by the French safety authorities. This examination is concerned with the remodelling of safety documents (safety reports, operation general regulations, incident and accident procedures, internal emergency plans and organization manuals for quality), with an analysis of the operating experience in frequently utilized circuits and with a systematic analysis of safeguards circuits. These evaluations are based on the experience of the Ardennes nuclear power plant, operating 15 years. This paper also describes the main studies and modifications undertaken in this power plant [fr

  12. Bid Preparation and Evaluation for Nuclear Power Plant Project Management

    International Nuclear Information System (INIS)

    Mohd Idris Taib, Mohd Khairulezwan Abdul Manan and Nur Farizan Amadzun

    2011-01-01

    Bid preparation and evaluation is one of the main activities in Nuclear Power Plant Project management. International Atomic Energy Agency guide and Korean experience was studied for Malaysian requirement in realization of first Nuclear Power Plant. Several aspects shall be taken into consideration such as political scenario, financial capabilities, sitting, human resource, technologies, fuel supplies and decommissioning for long term exceeded hundred years. Bidding process and activities is proposed for our country requirement. The main activities included but unlimited to Bid Invitation Specification, Bid Evaluation Process, Technical Evaluation, Economic Bid Evaluation and Contracting. On the end of day, Malaysia need safe and reliable Nuclear Power Plant. Malaysian Economic Transformation Programme also get benefit from spin-off localization products and services as well as Technology Transfer Programme. (author)

  13. Systematic safety evaluation of old nuclear power plants

    International Nuclear Information System (INIS)

    Dredemis, G.; Fourest, B.

    1984-01-01

    The French safety authorities have undertaken a systematic evaluation of the safety of old nuclear power plants. Apart from a complete revision of safety documents (safety analysis report, general operating rules, incident and accident procedures, internal emergency plan, quality organisation manual), this examination consisted of analysing the operating experience of systems frequently challenged and a systematic examination of the safety-related systems. This paper is based on an exercise at the Ardennes Nuclear Power Plant which has been in operation for 15 years. This paper also summarizes the main surveys and modifications relating to this power plant. (orig.)

  14. The evaluation of environmental effects of nuclear power plants

    International Nuclear Information System (INIS)

    Oezyurt, M.; Iyit, L.; Seyitogullari, S.

    2006-01-01

    Energy is today one of the most significant topics in the world. Humans are investigating alternative energy resources due to the fossil energy sources to be exhausted in future. As known, the life of energy resources such as coal and oil is limited. Natural gas will cover the need just for a limited period. On the other hand, developing population will increase the need of energy for the next generation. Therefore, alternative energy has gained much significance in recent years. Nuclear energy is the most criticized energy in public opinion. About 17 pct. of the electric need in the world is being covered by nuclear power plants . This ratio is over 30 pct. in European Union and over 78.2 pct. in France. The most significant risk as regard with environmental pollution is radioactive wastes for these plants. The opposite sides towards nuclear energy claim about the accidents of nuclear power plants and deaths in short and long terms. As long as the security rules are applied, nuclear power plants affect neither human nor environmental health in a detrimental way. The radiation emission scattered by nuclear power plants is very low. In this work, first of all nuclear energy was evaluated from a standpoint of environmental pollution and both positive and negative effects were investigated. As a result, the humanity will have to benefit from all the alternative energy resources , the nuclear energy as well, in order not to live in a dark world. Every technology has its own risks. It seems that if nuclear energy power plants are operated in high technology conditions it will be un given up for humanity

  15. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Kwon, J. D. [Yeungnam Univ., Gyeongsan (Korea, Republic of); Kang, K. J. [Chonnam National Univ., Gwangju (Korea, Republic of)] (and others)

    2001-03-15

    This research focuses on development of reliable life evaluation technology for nuclear power plant (NPP) components, and is divided into two parts, development of life evaluation systems for pressurized components and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered: development of expert systems for integrity assessment of pressurized components, development of integrity evaluation systems of steam generator tubes, prediction of failure probability for NPP components based on probabilistic fracture mechanics, development of fatigue damage evaluation technique for plant life extension, domestic round robin analysis for pressurized thermal shock of reactor vessels, domestic round robin analysis of constructing P--T limit curves for reactor vessels, and development of data base for integrity assessment. For evaluation of applicability of emerging technology to operating plants, on the other hand, the following eight topics are covered: applicability of the Leak-Before-Break analysis to Cast S/S piping, collection of aged material tensile and toughness data for aged Cast S/S piping, finite element analyses for load carrying capacity of corroded pipes, development of Risk-based ISI methodology for nuclear piping, collection of toughness data for integrity assessment of bi-metallic joints, applicability of the Master curve concept to reactor vessel integrity assessment, measurement of dynamic fracture toughness, and provision of information related to regulation and plant life extension issues.

  16. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim, Yun Jae; Choi, Jae Boong [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2002-03-15

    This project focuses on developing reliable life evaluation technology for nuclear power plant components, and is divided into two parts, development of a life evaluation system for nuclear pressure vessels and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered in this project: defect assessment method for steam generator tubes, development of fatigue monitoring system, assessment of corroded pipes, domestic round robin analysis for constructing P-T limit curve for RPV, development of probabilistic integrity assessment technique, effect of aging on strength of dissimilar welds, applicability of LBB to cast stainless steel, and development of probabilistic piping fracture mechanics.

  17. Dynamic testing of nuclear power plant structures: an evaluation

    International Nuclear Information System (INIS)

    Weaver, H.J.

    1980-02-01

    Lawrence Livermore Laboratory (LLL) evaluated the applications of system identification techniques to the dynamic testing of nuclear power plant structures and subsystems. These experimental techniques involve exciting a structure and measuring, digitizing, and processing the time-history motions that result. The data can be compared to parameters calculated using finite element or other models of the test systems to validate the model and to verify the seismic analysis. This report summarizes work in three main areas: (1) analytical qualification of a set of computer programs developed at LLL to extract model parameters from the time histories; (2) examination of the feasibility of safely exciting nuclear power plant structures and accurately recording the resulting time-history motions; (3) study of how the model parameters that are extracted from the data be used best to evaluate structural integrity and analyze nuclear power plants

  18. The evaluation of site characteristics for Guangdong nuclear power plant

    International Nuclear Information System (INIS)

    Zhou Ruming; Wu Dizhong; Yan Zhongmin

    1987-01-01

    This paper gives an account of the features of the site of Guangdong Nuclear Power Plant (GNPP) in general and in particular evaluates the outstanding site characteristics related to nuclear safety and public health. It is composed of two parts: the first part describes the seismo-geologic conditions of the site and the other treats the atmospheric dispersion conditions. It also contains the discussion why the possibility of inhabitancy within 5 km from the exclusion ares boundary would not be affected. (author)

  19. Seismic re-evaluation of French nuclear power plants

    International Nuclear Information System (INIS)

    Andrieu, R.

    1995-01-01

    After a presentation of the seismic inputs which have been taken into account in the design of the French Nuclear Power Plants, the re-assessed values of these inputs are shown. Some considerations about the specificity of the French PWR program with regard to the standardisation of plants are given together with the present objectives of seismic re-evaluations. Finally the main results of the seismic re-analysis being performed for the Phenix Fast Reactor are considered. (author)

  20. Evaluation and assessment of nuclear power plant seismic methodology

    Energy Technology Data Exchange (ETDEWEB)

    Bernreuter, D.; Tokarz, F.; Wight, L.; Smith, P.; Wells, J.; Barlow, R.

    1977-03-01

    The major emphasis of this study is to develop a methodology that can be used to assess the current methods used for assuring the seismic safety of nuclear power plants. The proposed methodology makes use of system-analysis techniques and Monte Carlo schemes. Also, in this study, we evaluate previous assessments of the current seismic-design methodology.

  1. Evaluation and assessment of nuclear power plant seismic methodology

    International Nuclear Information System (INIS)

    Bernreuter, D.; Tokarz, F.; Wight, L.; Smith, P.; Wells, J.; Barlow, R.

    1977-01-01

    The major emphasis of this study is to develop a methodology that can be used to assess the current methods used for assuring the seismic safety of nuclear power plants. The proposed methodology makes use of system-analysis techniques and Monte Carlo schemes. Also, in this study, we evaluate previous assessments of the current seismic-design methodology

  2. Technical evaluation of bids for nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    In continuation of its efforts to provide comprehensive and impartial guidance to Member States facing the need to introduce nuclear power, the International Atomic Energy Agency is issuing this guidebook as part of a series of guidebooks and codes of practice and, in particular, as a necessary supplement to 'Economic Evaluation of Bids for Nuclear Power Plants: A Guidebook', published by the IAEA in 1976 as Technical Reports Series No.175. The present publication is intended for project managers and senior engineers of electric utilities who are concerned with the evaluation of bids for a nuclear power project. It assumes that the reader has a good knowledge of the technical characteristics of nuclear power plants and of nuclear power project implementation. Its purpose is to provide the information necessary to organize, guide and supervise the technical evaluation of bids for a nuclear power project. It goes without saying that the technical staff carrying out the evaluation must have prior technical experience which cannot be provided by a guidebook

  3. Evaluation of Perry Nuclear Power Plant Unit 1 technical specifications

    International Nuclear Information System (INIS)

    Baxter, D.E.; Bruske, S.J.

    1985-11-01

    This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Perry Nuclear Power Plant Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Perry T/S. Several discrepancies were identified and subsequently resolved through telephone conversations with the staff reviewer and the utility representative. Pending completion of the resolutions noted in Parts 3 and 4 of this report, the Perry Nuclear Power Plant Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

  4. Emergy Evaluation of a Swedish Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kindberg, Anna

    2007-03-01

    Today it is common to evaluate and compare energy systems in terms of emission of greenhouse gases. However, energy systems should not only reduce their pollution but also give a large energy return. One method used to measure energy efficiency is emergy (embodied energy, energy memory) evaluation, which was developed by the system ecologist Howard T. Odum. Odum defines emergy as the available energy of one kind previously used up directly and indirectly to make a service or product. Both work of nature and work of human economy in generating products and services are calculated in terms of emergy. Work of nature takes the form of natural resources and work of human economy includes labour, services and products used to transform natural resources into something of value to the economy. The quotient between work of nature and work of human economy gives the emergy return on investment of the investigated product. With this in mind the present work is an attempt to make an emergy evaluation of a Swedish nuclear power plant to estimate its emergy return on investment. The emergy return on investment ratio of a Swedish nuclear power plant is calculated to approximately 11 in this diploma thesis. This means that for all emergy the Swedish economy has invested in the nuclear power plant it gets 11 times more emergy in return in the form of electricity generated by nuclear power. The method used in this work may facilitate future emergy evaluations of other energy systems

  5. Emergy Evaluation of a Swedish Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kindberg, Anna

    2007-03-15

    Today it is common to evaluate and compare energy systems in terms of emission of greenhouse gases. However, energy systems should not only reduce their pollution but also give a large energy return. One method used to measure energy efficiency is emergy (embodied energy, energy memory) evaluation, which was developed by the system ecologist Howard T. Odum. Odum defines emergy as the available energy of one kind previously used up directly and indirectly to make a service or product. Both work of nature and work of human economy in generating products and services are calculated in terms of emergy. Work of nature takes the form of natural resources and work of human economy includes labour, services and products used to transform natural resources into something of value to the economy. The quotient between work of nature and work of human economy gives the emergy return on investment of the investigated product. With this in mind the present work is an attempt to make an emergy evaluation of a Swedish nuclear power plant to estimate its emergy return on investment. The emergy return on investment ratio of a Swedish nuclear power plant is calculated to approximately 11 in this diploma thesis. This means that for all emergy the Swedish economy has invested in the nuclear power plant it gets 11 times more emergy in return in the form of electricity generated by nuclear power. The method used in this work may facilitate future emergy evaluations of other energy systems.

  6. Evaluation of human error estimation for nuclear power plants

    International Nuclear Information System (INIS)

    Haney, L.N.; Blackman, H.S.

    1987-01-01

    The dominant risk for severe accident occurrence in nuclear power plants (NPPs) is human error. The US Nuclear Regulatory Commission (NRC) sponsored an evaluation of Human Reliability Analysis (HRA) techniques for estimation of human error in NPPs. Twenty HRA techniques identified by a literature search were evaluated with criteria sets designed for that purpose and categorized. Data were collected at a commercial NPP with operators responding in walkthroughs of four severe accident scenarios and full scope simulator runs. Results suggest a need for refinement and validation of the techniques. 19 refs

  7. Evaluation of nuclear plant cable aging through condition monitoring

    International Nuclear Information System (INIS)

    Extending the lifetime of a nuclear power plant [(hereafter referred to simply as NPP)] is one of the most important concerns in the global nuclear industry. Cables are one of the long - life items that have not been considered for replacement during the design life of a NPP. To extend the cable life beyond the design life, it is first necessary to prove that the design life is too conservative compared with actual aging. Condition monitoring is useful means of evaluating the aging condition of cable. In order to simulate natural aging in a nuclear power plant, a study on accelerated aging must first be conducted. In this paper, evaluations of mechanical aging degradation for a neoprene cable jacket were performed after accelerated aging under continuous and intermittent heating conditions. Contrary to general expectations, intermittent heating to the neoprene cable jacket showed low aging degradation, 50% break-elongation, and 60% indenter modulus, compared with continuous heating. With a plant maintenance period of 1 month after every 12 or 18 months operation, we can easily deduce that the life time of the cable jacket of neoprene can be extended much longer than estimated through the general EQ test, which adopts continuous accelerated aging for determining cable life. Therefore, a systematic approach that considers the actual environment conditions of the nuclear power plant is required for determining cable life

  8. Chemistry evaluation in French EDF Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jacquier, Hervé

    2014-01-01

    The Nuclear Production Division of EDF is comprised of 19 power stations (58 PWR reactors) and 2 national engineering organisations. Nuclear Inspection (IN) is an internal assessment unit of the EDF Nuclear Production Directorate. At the request of the Directorate, it carries out periodic evaluations of all the units of the division. The evaluation of the nuclear sites (EGE: Overall Excellence Assessment) is carried out every 4 years, an intermediate evaluation is also carried out between each EGE. These evaluations are independent of the WANO and IAEA evaluations. Exchanges are carried out between Nuclear Inspection and the other international operators (for example, USA (INPO), England, China...) to share site evaluation methods. These evaluations are carried out by a team of 30 inspectors, reinforced during each evaluation by 10 peers who come from the various French nuclear sites. Nuclear Inspection produces a performance standards document for each FUNCTIONAL AREA, which is based on the requirements of the company. On the whole, 13 areas are evaluated during each inspection, in particular: Management, Operations, Maintenance, Engineering and Chemistry. The area of reactor plant chemistry has been evaluated since 2009. The Chemistry performance standards document is written from the EDF internal requirements and international references. During site evaluations, all the performance standards are assessed for compliance. The Chemistry performance standards document is comprised of 3 topics: Management of plant chemistry, The respect of the chemical and radiochemical specifications, The condition of the laboratories and the sampling lines, measuring equipment, and chemical products. The evaluations carried out make it possible to define strengths and weaknesses which the sites must address. After each evaluation, the assessment is presented to the site management and to the director of EDF Nuclear Production. For 4 years these evaluations have allowed progress to

  9. Evaluation of masonry wall design at nuclear power plants

    International Nuclear Information System (INIS)

    Con, V.N.; Subramonian, N.; Chokshi, N.

    1983-01-01

    The structural integrity of safety-related masonry walls in operating nuclear power plants may not be maintained when subjected to certain loads and load combinations. The paper presents some findings based upon the review of the design and analysis procedures used by the licensees in the reevaluation of safety-related masonry walls. The design criteria developed by the Structural Engineering Branch (SEB) of the United States Nuclear Regulatory Commission (NRC) along with other standard codes such as the Uniform Building Code, ACI 531-79, ATC 3-06, and NCMA were used as guidance in evaluating the design criteria developed by the licensees. The paper deals with the following subject areas: loads and load combinations, allowable stresses, analytical procedures, and modification methods. The paper concludes that, in general, the masonry walls in nuclear power plants comply with the working stress design requirements. In some cases, certain nonlinear analysis methods were used. The applicability of these methods is discussed. (orig.)

  10. A methodology for nuclear power plant operational events evaluation

    International Nuclear Information System (INIS)

    Araujo, Jeferson

    2015-01-01

    Operational events are normal occurrences in industrial plants and in nuclear power plants. The evaluation of operational events gains importance when it comes specifically to nuclear power plants due to the proportions that the impact and the consequences of these events may cause to the installation itself, their workers, the external area of the nuclear installation, the environment and to the public in general. These consequences, for the operation of these facilities can range from very little, until the consequences that lead to accidents and can cause significant impacts. Operational events may be associated or have influence in many fields of knowledge, such as operation, maintenance, engineering, Radiological Protection, physical protection, chemistry, Human or Organizational Factors and external events, among others. The accident at the Fukushima Daichi nuclear power plant, shows the importance of exhausting all the studies concerning operational events in order to improve the operational safety of nuclear plants, considering all the causes and possible consequences. In this context, the evaluation of operational events discipline emerges as an important and relevant tool to contribute to the maintenance and/or improvement of the operational safety of nuclear installations. Not without reason the nuclear industry actively participates in programs of exchange of operational experience, where relevant events are thoroughly evaluated and discussed in specific forums, such as power plant operators, regulators and/or joint technical meetings, always with the purpose to prevent, minimize or mitigate its consequences. Any evaluation of operational events passes necessarily by an in-depth study of the circumstances of the event, culminating with the identification of your cause and proposition of corrective actions to prevent recurrence of similar events. Additionally, the events should not be studied individually, but evaluated within a temporal context in order

  11. Evaluation of the feasibility, economic impact, and effectiveness of underground nuclear power plants. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    Information on underground nuclear power plants is presented concerning underground nuclear power plant concepts; public health impacts; technical feasibility of underground concepts; economic impacts of underground construction; and evaluation of related issues.

  12. Evaluation of the feasibility, economic impact, and effectiveness of underground nuclear power plants. Final technical report

    International Nuclear Information System (INIS)

    1978-05-01

    Information on underground nuclear power plants is presented concerning underground nuclear power plant concepts; public health impacts; technical feasibility of underground concepts; economic impacts of underground construction; and evaluation of related issues

  13. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    International Nuclear Information System (INIS)

    Demick, L.E.

    2011-01-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

  14. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2011-10-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

  15. Destructiveness criteria for seismic risk evaluation of nuclear power plant

    International Nuclear Information System (INIS)

    Saragoni, G.R.

    1995-01-01

    Two criteria of destructiveness for seismic risk evaluation of nuclear power plant are presented. The first one is a simple linear criterion that allows to compute average response spectra in terms of earthquake accelerogram characteristics. The second defines the destructiveness potential factor P D which measures the capacity of earthquake to produce nonlinear damage. This second criterion that shows large differences of destructiveness capacity for earthquake accelerograms of different seismic environment, specially between subductive and transcursive, is strongly recommended. (author). 8 refs., 1 fig. 1 tab

  16. Economic evaluation of bids for nuclear power plants. 1999 edition

    International Nuclear Information System (INIS)

    2000-01-01

    The introduction of a nuclear power plant (NPP) in a country is a major undertaking for all entities involved. The necessary planning work and co-ordination of the different fields of interest, from the point of view of governments, authorities, industries, universities and of the general public, must be done on a long term basis. This IAEA report may help support the work of the utility in the bidding process, especially in the economic bid evaluation. The different methods, aspects and parameters described should be regarded as a guide. The report's target criterion of lowest levelised discounted electricity generation costs (LDEGC) is a very useful and practical way of ranking bids. In view of the huge investment needed, the owner's country must be fully committed to a nuclear programme. A nuclear programme also requires a guarantee of long term financing, which implies the provision of local and foreign contributions. The necessary staff for all of the various areas of a nuclear programme must be recruited and trained. Within the planning phase, the bid invitation specification (BIS) has to be prepared and sent out in order to receive bids for the scope of supply and services desired by the owner. The evaluation of the bids received from the suppliers in response to the BIS is a huge and long-term task. The evaluation process should lead to the selection of the best bidder and at least to the final decision on the partners constructing the NPP. The responsibility for the entire bidding process lies with the plant owner. For technology transfer, two requirements need to be satisfied: the owner needs a well established and experienced engineering capability, and the supplier must be ready to transfer the agreed technology in such a way as to support the project goals. During the bid evaluation process, all aspects of the technical, financial and contractual approaches must be considered. Nowadays, political, socioeconomic and public acceptance aspects play a

  17. Modelling operator cognitive interactions in nuclear power plant safety evaluation

    International Nuclear Information System (INIS)

    Senders, J.W.; Moray, N.; Smiley, A.; Sellen, A.

    1985-08-01

    The overall objectives of the study were to review methods which are applicable to the analysis of control room operator cognitive interactions in nuclear plant safety evaluations and to indicate where future research effort in this area should be directed. This report is based on an exhaustive search and review of the literature on NPP (Nuclear Power Plant) operator error, human error, human cognitive function, and on human performance. A number of methods which have been proposed for the estimation of data for probabilistic risk analysis have been examined and have been found wanting. None addresses the problem of diagnosis error per se. Virtually all are concerned with the more easily detected and identified errors of action. None addresses underlying cause and mechanism. It is these mechanisms which must be understood if diagnosis errors and other cognitive errors are to be controlled and predicted. We have attempted to overcome the deficiencies of earlier work and have constructed a model/taxonomy, EXHUME, which we consider to be exhaustive. This construct has proved to be fruitful in organizing our thinking about the kinds of error that can occur and the nature of self-correcting mechanisms, and has guided our thinking in suggesting a research program which can provide the data needed for quantification of cognitive error rates and of the effects of mitigating efforts. In addition a preliminary outline of EMBED, a causal model of error, is given based on general behavioural research into perception, attention, memory, and decision making. 184 refs

  18. Air toxics evaluation for a nuclear power plant

    International Nuclear Information System (INIS)

    Lokey, D.; Orr, W.

    1994-01-01

    An emission inventory of nonradiological hazardous air pollutants (HAP) was prepared for TVA's Browns Ferry Nuclear Plant (BFN). The purpose of this effort was to determine whether BFN is a major HAP emission source. About 40 specific HAP were identified in products used at FBN. HAP emission totals from BFN were estimated at 6 T/Y, well below the major-source cutoff. Off-site waste disposal reduced gross emission estimates by about 8 percent. HAP emission evaluation by product categories showed the largest portion due to paints and coatings, followed by cleaning compounds. HAP emission breakdown by chemical species showed xylene to be the predominate HAP emitted, followed by methyl ethyl ketone

  19. An evaluation of water hammer in nuclear power plants

    International Nuclear Information System (INIS)

    Serkiz, A.

    1983-01-01

    Water hammer can occur as a result of to steam-void collapse, steam-driven slugs of water, pump startup into voided lines, and inadverdent closures of valves; the result is large hydraulic pressure loads. Since 1968, about 150 water hammer events have been reported in U.S. nuclear power plants; damage has been confined to pipe hangers and snubbers. The increase in reported water hammer events in the 1970s led to the designation of water hammer as Unresolved Safety Issue A-1 in 1979. Since the mid-1970s reported occurrences and their underlying phenomena have been studied, and corrective design and operational measures have been implemented. This paper presents the results of generic evaluations of water hammer occurrence and reported damage and sets forth technical findings. These findings indicate that good design practices and operational procedures can minimize the occurrence of water hammer

  20. Safety evaluation of nuclear power plant against the virtual tsunami

    International Nuclear Information System (INIS)

    Chin, S. B.; Imamura, Fumihiko

    2004-01-01

    The main scope of this study is the numerical analysis of virtual tsunami event near the Ulchin Nuclear Power Plants. In the numerical analysis, the maximum run-up height and draw-down are estimated at the Ulchin Nuclear Power Plants. The computer program developed in this study describes the propagation and associated run-up process of tsunamis by solving linear and nonlinear shallow-water equations with finite difference methods. It can be used to check the safety of a nuclear power plant against tsunami attacks. The program can also be used to calculate run-up height of wave and provide proper design criteria for coastal facilities and structures. A maximum inundation zone along the coastline can be developed by using the moving boundary condition. As a result, it is predicted that the Ulchin Nuclear Power Plants might be safe against the virtual tsunami event. Although the Ulchin Nuclear Power Plants are safe against the virtual tsunami event, the occurrence of a huge tsunami in the seismic gap should be investigated in detail. Furthermore, the possibility of nearshore tsunamis around the Korean Peninsula should also be studied and monitored continuously

  1. A re-evaluation of nuclear plant offsite power supplies

    International Nuclear Information System (INIS)

    William E Berger; Robert E Henry

    2005-01-01

    Full text of publication follows: De-regulation of the electric power industry has resulted in separate ownership of the transmission and power generation facilities as well as a revised format for operating the transmission facilities. Currently we see the transfer of large blocks of bulk power between markets which can impact the voltage regulation at the offsite power supply. Where Nuclear Plant operations once knew with a large degree of certainty the operating range of the system supplying the offsite power supply, this may no longer be the case and more challenges to the safety systems could result. These challenges may manifest themselves as either a loss of offsite power or voltage levels approaching the degraded level setpoints. In this paper we will first explore what challenges are caused by deregulation and how they impact offsite power supply operations. Next we will incorporate the knowledge grained regarding accidents and consequences from the Individual Plant Evaluations (IPE's) to see how the offsite power supply could be operated to mitigate the challenges and extend the capacity of the auxiliary power system. Various scenarios will be examined using the Modular Accident Analysis Program (MAAP) as an integral plant model. MAAP simulations that include both the plant thermal hydraulic responses and corresponding electric power demand are presented to demonstrate the impact of alternate approaches to offsite power system operation. The original design phase of the offsite and onsite power distribution system was based on a criterion relating to the starting of all safety loads if a safety injection signal was present independent of the accident or its progression. The IPE and risk informed insights that are readily available today will be applied in the re-analyses of the offsite distribution system response. (authors)

  2. 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

  3. Safety evaluation of the Laguna Verde nuclear power plant

    International Nuclear Information System (INIS)

    Delgado G, J.L.

    1991-01-01

    The present work describe the licensing process for the first nuclear power plant built in Mexico, it presents the difficulties found during the several years of construction and tests until the phrase a level of safety equivalent to that of the country of origin of the nuclear steam supply system could be applicable to Laguna Verde, at least from the point of view of the mexican regulatory body, and also that this statement could be signed for the inspectors of international organizations. (author)

  4. A methodology for evaluating ''new'' technologies in nuclear power plants

    International Nuclear Information System (INIS)

    Korsah, K.; Clark, R.L.; Holcomb, D.E.

    1994-01-01

    As obsolescence and spare parts issues drive nuclear power plants to upgrade with new technology (such as optical fiber communication systems), the ability of the new technology to withstand stressors present where it is installed needs to be determined. In particular, new standards may be required to address qualification criteria and their application to the nuclear power plants of tomorrow. This paper discusses the failure modes and age-related degradation mechanisms of fiber optic communication systems, and suggests a methodology for identifying when accelerated aging should be performed during qualification testing

  5. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae [Sungkwunkwan Univ., Seoul (Korea, Republic of); Lee, Hyang Beom [Soongsil Univ., Seoul (Korea, Republic of); Shin, Young Kil [Kunsan National Univ., Gunsan (Korea, Republic of); Chung, Hyun Jo [Wonkwang Univ., Iksan (Korea, Republic of); Park, Ik Keun; Park, Eun Soo [Seoul National University of Technology, Seoul (Korea, Republic of)

    2001-03-15

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing.

  6. Development of life evaluation technology for nuclear power plant components

    International Nuclear Information System (INIS)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae; Lee, Hyang Beom; Shin, Young Kil; Chung, Hyun Jo; Park, Ik Keun; Park, Eun Soo

    2001-03-01

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing

  7. Seismic evaluation and strengthening of Bohunice nuclear power plant structures

    International Nuclear Information System (INIS)

    Shipp, J.G.; Short, S.A.; Grief, T.; Borov, V.; Kuzma, J.

    2001-01-01

    A seismic assessment and strengthening investigation is being performed for selected structures at the Bohunice V1 Nuclear Power Plant in Slovakia. Structures covered in this paper include the reactor building complex and the emergency generator station. The emergency generator station is emphasized in the paper as work is nearly complete while work on the reactor building complex is ongoing at this time. Seismic evaluation and strengthening work is being performed by a cooperative effort of Siemens and EQE along with local contractors. Seismic input is the interim Review Level Earthquake (horizontal peak ground acceleration of 0.3 g). The Bohunice V1 reactor building complex is a WWER 4401230 nuclear power plant that was originally built in the mid-1970s but had extensive seismic upgrades in 1991. Siemens has performed three dimensional dynamic analyses of the reactor building complex to develop seismic demand in structural elements. EQE is assessing seismic capacities of structural elements and developing strengthening schemes, where needed. Based on recent seismic response analyses for the interim Review Level Earthquake which account for soil-structure interaction in a rigorous manner, the 1991 seismic upgrade has been found to be inadequate in both member/connection strength and in providing complete load paths to the foundation. Additional strengthening is being developed. The emergency generator station was built in the 1970s and is a two-story unreinforced brick masonry (URM) shear wall building above grade with a one story reinforced concrete shear wall basement below grade. Seismic analyses and testing of the URM walls has been performed to assess the need for building strengthening. Required structural strengthening for in-plane forces consists of revised and additional vertical steel framing and connections, stiffening of horizontal roof bracing, and steel connections between the roof and supporting walls and pointing of two interior transverse URM

  8. Nuclear versus fossil power plants: evolution of economic evaluation techniques

    International Nuclear Information System (INIS)

    Thuesen, G.J.

    1975-01-01

    The purpose of this presentation is to document the evolution of methods used by an electric utility for comparing the economic attractiveness of nuclear versus fossil electric power generation. This process of change is examined as it took place within the Georgia Power Company (GPC), a company spending in the neighborhood of half a billion dollars annually for capital improvements. This study provides a look at the variety of richness of information that can be made available through the application of different methods of economic analysis. In addition, the varied presentations used to disclose relationships between alternatives furnish evidence as to the effectiveness of providing pertinent information in a simple, meaningful manner. It had been generally accepted throughout GPC that nuclear power was economically desirable as an alternative for the production of base-load power. With inflation increasing, its advantage over fossil power appeared to be significantly increasing as the large operating costs of fossil generation seemed to be more vulnerable to inflation than the costs of operating a nuclear facility. An early indication that the company should reevaluate this position was the experience gained with the installation of their first nuclear plant. Here, actual total costs were exceeding their original construction estimates by a factor of two. Thus the question arose ''Does the high capital cost of nuclear generation offset its operating advantages when compared to similar-sized coal-burning plants.'' To answer this question, additional analyses were undertaken

  9. Safety evaluation of the nuclear power plant at Cattenom

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This is a systematic compilation of the material which was dealt with at the level of the German-French Commission (on questions of the safety of nuclear installations) in this discussions about the nuclear power plant at Cattenom. As a supplement to the report published already in 1982, the Commission has officially released its deliberation results on the subjects constructive safety measures, radiological effects, and precautions in case of an emergency. The allegations according to which the installation is wanting in safety are countered by the joint statement of the chairmen of GPR (Permanent Group on Reactors) and RSK (German Commission on Reactor Safety) of August 29, 1986. (HSCH) [de

  10. Nuclear power plant personnel training and its evaluation. A guidebook

    International Nuclear Information System (INIS)

    1996-01-01

    The Guidebook will prove especially useful for, and is addressed primarily to: nuclear power operating organizations establishing or upgrading their NPP personnel training systems; regulatory personnel responsible for setting requirements and/or evaluating NPP personnel training; and organizations (within or outside the operating organization) responsible for the development, implementation and evaluation of training programmes for NPP personnel. Figs, tabs

  11. An evaluation method of fault-tolerance for digital plant protection system in nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Jun Seok; Kim, Man Cheol; Seong, Poong Hyun; Kang, Hyun Gook; Jang, Seung Cheol

    2005-01-01

    In recent years, analog based nuclear power plant (NPP) safety related instrumentation and control (I and C) systems have been replaced to modern digital based I and C systems. NPP safety related I and C systems require very high design reliability compare to the conventional digital systems so that reliability assessment is very important. In the reliability assessment of the digital system, fault tolerance evaluation is one of the crucial factors. However, the evaluation is very difficult because the digital system in NPP is very complex. In this paper, the simulation based fault injection technique on simplified processor is used to evaluate the fault-tolerance of the digital plant protection system (DPPS) with high efficiency with low cost

  12. An economic evaluation of nuclear power plant externalities, using the hedonic price approach

    OpenAIRE

    Kato, Hirotaka; Ueta, Kazuhiro

    2012-01-01

    This study evaluates the external costs of nuclear power plants in Japan. Using the hedonic price approach, we analyze changes in land price in Fukui Prefecture before and after the Fukushima nuclear accident. The land in question is located far from the Fukushima accident area and was not directly impacted by the accident itself. The results of this study reveal that the area affected by the nuclear power plant's negative externalities expanded following the accident, and that the intensity ...

  13. 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

  14. Evaluation of environmental data relating to selected nuclear power plant sites. Prairie Island Nuclear Generating Plant site

    International Nuclear Information System (INIS)

    Murarka, I.P.

    1976-11-01

    Environmental monitoring data for 1973 through 1975 pertaining to the Prairie Island Nuclear Generating Station (which began commercial operation in December 1973) were analyzed by the most practical qualitative and quantitative methods. Evaluations of aquatic and terrestrial biotic data are presented in this report. The data indicate no significant immediate deleterious effects on the biota from plant operation, thus confirming preoperational predictions. Although the station has not operated long enough to reveal long-term deleterious effects, present indications do not lead to a concerned prediction that any are developing. Recommendations are suggested for improving monitoring techniques

  15. Evaluation of nuclear power plant environmental impact prediction, based on monitoring programs. Summary and recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Gore, K.L.; Thomas, J.M.; Kannberg, L.D.; Watson, D.G.

    1977-02-01

    An evaluation of the effectivenss of non-radiological environmental monitoring programs is presented. The monitoring programs for Monticello, Haddam Neck, and Millstone Nuclear Generating Plants are discussed. Recommendations for improvements in monitoring programs are presented.

  16. Computerized planning system for nuclear power plant evaluation

    International Nuclear Information System (INIS)

    Bonczek, R.H.; Holsapple, C.W.; Whinston, A.B.

    1976-01-01

    A computerized system is described for information storage and query processing adapted to complex socio-technological issues. The system is referred to as GPLAN (Generalized Planning System) and can accommodate both qualitative (verbal) and quantitative data. The issue illustrated is the construction of a nuclear power plant, and involves interdisciplinary research and planning. The system's outstanding features are the use of the network variety of data base, the selective retrieval of any configuration of data from a particular network structure, automatic execution of any desired application program from a standard or special library of applications, user interface with a data base and applications by submitting English-like, non-procedural queries, and generality which allows tailoring to specific applications and provides a basis for integration of planning and research activities. The system is general and can be used for a wide variety of socio-technological issues which involve complex data relationships

  17. Integrated Plant Safety Assessment, Systematic Evaluation Program: Yankee Nuclear Power Station (Docket No. 50-29)

    International Nuclear Information System (INIS)

    1987-10-01

    The US Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0825), under the scope of the Systematic Evaluation Program (SEP), for Yankee Atomic Electric Company's Yankee Nuclear Power Station located in Rowe, Massachusetts. The SEP was initiated by the NRC to review the design of older operating nuclear power plants to reconfirm and document their safety. This report documents the review completed under the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations after the Final IPSAR for the Yankee plant was issued. The review has provided for (1) an assessment of the significance of differences between current technical positions on selected safety issues and those that existed when Yankee was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. 2 tabs

  18. Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Abbott, L.

    1985-09-01

    An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures

  19. Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L [ed.

    1985-09-01

    An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures.

  20. 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

  1. An evaluation of the dismantling technologies for decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Jeong, KwanSeong; Choi, ByungSeon; Moon, Jeikwon; Hyun, Dongjun; Lee, JongHwan; Kim, IkJune; Kim, GeunHo; Seo, JaeSeok

    2014-01-01

    Highlights: • This paper is evaluation method on the technologies for decommissioning of nuclear power plants. • The parameters of evaluation are performance, site-specific, safety, and cost impacts. • The evaluation model was applied for dismantling of a steam generator. - Abstract: This paper is to suggest an evaluation method on the dismantling technologies for decommissioning of nuclear power plants. The parameters of evaluation are performance impacts, site-specific impacts, safety impacts, and cost impacts. The evaluation model was provided and applied for dismantling of a steam generator

  2. Evaluation of emergency operating procedures for nuclear power plants. Final report Feb 80-Mar 81

    International Nuclear Information System (INIS)

    Fuchs, F.; Engelschall, J.; Imlay, G.

    1981-04-01

    The report documents the first of two tasks performed for the NRC relative to emergency operating procedures for nuclear power plants with special emphasis on LOCA procedures. Emergency procedures from nine plants have been evaluated. The recommendations regarding each plant's emergency procedures were made in separate reports. In this report, an analysis is made of problems common to emergency procedures in all nine plants. Recommendations are advanced for improvement in presentation style, level of detail, and administrative control

  3. Guidelines for the technical evaluation of replacement items in nuclear power plants (NCIG-11)

    International Nuclear Information System (INIS)

    Craig, W.E.; Fakhar, A.A.; Shulman, M.N.

    1989-12-01

    This document presents guidelines and supporting information for the technical evaluation of replacement items in nuclear power plants. These guidelines contain six major sections which provide the practical knowledge and a programmatic approach to determine the technical and quality requirements necessary to generate purchase documents to procure the proper replacement items. The technical evaluation methodology includes the following steps. (1) Identification of the need for a technical evaluation. (2) Component/part functional classification procedures. (3) Performance of a Failure Modes and Effects Analysis. (4) Selection of Critical Characteristics for Design Determination. (5) Performance of a ''Like-For-Like'' or ''Alternate'' item Evaluation. (6) Preparation of the Technical and Quality Requirements Specification. Work on this document was initiated in response to the increased emphasis by the utilities owning nuclear power plants and nuclear industry on procurement of replacement items for use in safety related applications at nuclear power plants. 20 refs., 9 figs., 14 tabs

  4. Site selection and evaluation for nuclear power plants with respect to population distribution

    International Nuclear Information System (INIS)

    1980-01-01

    This safety guide, relating population distribution to site selection and evaluation, for nuclear power plants, forms part of the IAEA's programme, referred to as the NUSS programme (Nuclear Safety Standards). The guide presents population distribution data, requirements, examples of site screening methods, and an overview of radiological impact assessment with respect to population distribution

  5. Manual on quality assurance for the survey, evaluation and confirmation of nuclear power plant sites

    International Nuclear Information System (INIS)

    1987-04-01

    The present Manual on Quality Assurance for the Survey, Evaluation and Confirmation of Nuclear Power Plant Sites contains supporting material and illustrates examples for implementing the requirements contained in the Code of Practice on Quality Assurance for Safety in Nuclear Power Plants to the activities of survey, evaluation and confirmation of nuclear power plant sites. At the same time the Code of Practice for Safety in Nuclear Power Plant Siting, and Safety Guides in the siting series contain requirements and recommendations to implement a quality assurance programme in selected activities of the siting process. This manual is intended to provide guidance and illustrate examples on this implementation. During preparation and reviews of this Manual it was found out that the methodology of implementation of the quality assurance programme in siting activities is still under development. For these reasons it was considered appropriate to publish this Manual as a temporary publication for trial use

  6. Virtual reality applied in the ergonomic evaluation of nuclear power plant control room

    International Nuclear Information System (INIS)

    Gatto, Leandro Barbosa da Silveira

    2012-01-01

    A nuclear power plant control room is a complex system that controls a nuclear and thermodynamic process used to produce electrical energy. The operators interact with the control room through interfaces that have significant implications to nuclear power plant safety and influence the operator activity. The operator activity presents complexity features and shows a series of mechanisms absents from the human factors guidelines, important to the evaluation and update of control rooms. The ergonomics approach considers the operation strategies, the interaction between the operators, the operator-system interaction, and interaction between operators and support groups. The main objective of this paper is propose the modeling of a nuclear control room, with the support of a game engine core. This tool will be used in the ergonomic evaluation of nuclear control room, generating information and data that will make possible the adequacy of control rooms features to the legal requirements of the regulating agency, assisting the nuclear licensing. (author)

  7. Development of resilience evaluation method for nuclear power plant. Part 1. Proposal of resilience index for assessment of safety of nuclear power plant under severe accident

    International Nuclear Information System (INIS)

    Demachi, Kazuyuki; Suzuki, Masaaki; Itoi, Tatsuya

    2016-01-01

    In this research, a new index 'The Resilience Index' was proposed to evaluate the capability of nuclear power plant to recover from the situation of safety function lost. Three elements assumed to evaluate the resilience index are the achievement rate, necessary time, and probability of success of each accident management activity. The resilience index is expected to visualize the improvement of safety of each nuclear power plant against severe accidents. (author)

  8. Nuclear plant power up-rate study: feedwater heater evaluations

    International Nuclear Information System (INIS)

    Svensson, Eric; Catapano, Michael; Coakley, Michael; Thomas, Dan

    2014-01-01

    Given today's nuclear industry business climate, it has become common for Utility companies to consider increasing unit capacities through turbine replacement and power up-rates. An integral part of the studies conducted by many towards this end, involve the generation of a set of turbine cycle heat balances with predicted performance parameters for the up-rated condition. Once these tentative operating values are established, it becomes necessary to evaluate the suitability of the existing components within each system to ensure they are capable of continued safe and reliable operation. The ultimate cost for the up-rate, including the cost for any major required modifications or significant replacements is weighed against increased revenue generated from the up-rate over time. Exelon's Peach Bottom Atomic Power Station (PBAPS) is currently planning for an Extended Power up-rate (EPU) for both units. To ensure the existing Feedwater Heaters (FWH) could maintain the operating and transient response margins at the EPU condition, an engineering study was conducted. Powerfect Inc. in conjunction with SPX Heat Transfer LLC were contracted to provide engineering services to analyze the design, thermal performance, reliability and operating conditions at projected EPU conditions. Specifically, to address the following with regard to the station's Feedwater Heaters (FWHs): 1. Evaluate Drain Cooler (DC) Velocities - including zone inlet velocity, cross and window velocities and outlet velocities. 2. Evaluate Drain Cooler Zone Pressure Drop for effect on drain cooler margins to flashing. 3. Evaluate differential pressure allowable across the pass partition plate. 4. Evaluate Drain Cooler Tube Vibration Potential. 5. Perform detailed steam dome velocity calculations. The goal of the study was to identify the most susceptible areas within the heaters for problems and potential failures when operating at the higher duty of the EPU condition for the remaining life

  9. Probabilistic safety assessment technology for commercial nuclear power plant security evaluation

    International Nuclear Information System (INIS)

    Liming, J.K.; Johnson, D.H.; Dykes, A.A.

    2004-01-01

    Commercial nuclear power plant physical security has received much more intensive treatment and regulatory attention since September 11, 2001. In light of advancements made by the nuclear power industry in the field of probabilistic safety assessment (PSA) for its power plants over that last 30 years, and given the many examples of successful applications of risk-informed regulation at U. S. nuclear power plants during recent years, it may well be advisable to apply a 'risk-informed' approach to security management at nuclear power plants from now into the future. In fact, plant PSAs developed in response to NRC Generic Letter 88-20 and related requirements are used to help define target sets of critical plant safety equipment in our current security exercises for the industry. With reasonable refinements, plant PSAs can be used to identify, analyze, and evaluate reasonable and prudent approaches to address security issues and associated defensive strategies at nuclear power plants. PSA is the ultimate scenario-based approach to risk assessment, and thus provides a most powerful tool in identifying and evaluating potential risk management decisions. This paper provides a summary of observations of factors that are influencing or could influence cost-effective or 'cost-reasonable' security management decision-making in the current political environment, and provides recommendations for the application of PSA tools and techniques to the nuclear power plant operational safety response exercise process. The paper presents a proposed framework for nuclear power plant probabilistic terrorist risk assessment that applies these tools and techniques. (authors)

  10. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  11. On the economic evaluation of bids for construction of nuclear power plants

    International Nuclear Information System (INIS)

    El Osery, I.A.

    1988-01-01

    The work presented is a part of R and D program conducted by nuclear power plants authority and aiming at the development of the tools necessary for economic evaluation of bids for construction of nuclear power plants in Egypt. The paper describes the potential of economic evaluation of bids and the methods generally used for such purpose. The mathematical tools proposed for the calculation of the different cost items; involved in the economic evaluation, are presented. Account for expected changes in the economic environment is considered including exchange rate of foreign currency, price escalation, and inflation rates. The proposed model takes into consideration the financing terms and conditions. The proposed approach has already been incorporated in a computer program called ''ECONEV'' developed by the author for evaluation of the international bids for construction of the first nuclear power plant in Egypt at El-Dabaa site (160 Km west of Alexandria)

  12. XOQDOQ, Meteorological Evaluation of Atmospheric Nuclear Power Plant Effluents

    International Nuclear Information System (INIS)

    Sagendorf, J.F.; Goll, J.T.; Sandusky, W.F.; Eyberger, L.R.

    1990-01-01

    1 - Description of program or function: XOQDOQ was designed for meteorological evaluation of continuous and anticipated intermittent releases from commercial nuclear power reactors. It calculates annual relative effluent concentrations and average relative deposition values at locations specified by the user and at various standard radial distances and segments for downwind sectors. It also calculates these values at the specified locations for anticipated intermittent (e.g. containment or purge) releases, which occur during routine operation. The program computes an effective plume height that accounts for physical release height, aerodynamic down-wash, plume rise, and terrain features. The user may optionally select additional plume dispersion due to building wakes, plume depletion via dry deposition, and plume radioactive decay, or specify adjustments to represent non-straight line trajectories (recirculation or stagnation). 2 - Method of solution: XOQDOQ is based on the principle that diffusion of material released to the atmosphere can be described by a Gaussian distribution within the plume with transport described by a straight-line trajectory. The horizontal and vertical dispersion coefficients are empirically determined, largely from observations at or near ground level. The program implements the assumptions outlined in Section C of NRC Regulatory Guide 1.111. Long-term average values of relative effluent concentration are calculated by assuming a long-term continuous release with effluent distributed evenly across a 22-1/2 degree sector. 3 - Restrictions on the complexity of the problem - Maxima of: 30 receptor locations/receptor type, 14 wind-speed classes, 10 distances of site-specific recirculation correction factors, 8 receptor types, 7 atmospheric stability categories, 5 separate release points. XOQDOQ cannot handle multiple emission sources or plume depletion via wet deposition, or evaluate the meteorological aspects of the consequences of

  13. Operational safety performance and economical efficiency evaluation for nuclear power plants

    International Nuclear Information System (INIS)

    Liu Yachun; Zou Shuliang

    2012-01-01

    The economical efficiency of nuclear power includes a series of environmental parameters, for example, cleanliness. Nuclear security is the precondition and guarantee for its economy, and both are the direct embodiment of the social benefits of nuclear power. Through analyzing the supervision and management system on the effective operation of nuclear power plants, which has been put forward by the International Atomic Energy Agency (IAEA), the World Association of Nuclear Operators (WANO), the U.S. Nuclear Regulatory Commission (NRC), and other organizations, a set of indexs on the safety performance and economical efficiency of nuclear power are explored and established; Based on data envelopment analysis, a DEA approach is employed to evaluate the efficiency of the operation performance of several nuclear power plants, Some primary conclusion are achieved on the basis of analyzing the threshold parameter's sensitivity and relativity which affected operational performance. To address the conflicts between certain security and economical indicators, a multi-objective programming model is established, where top priority is given to nuclear safety, and the investment behavior of nuclear power plant is thereby optimized. (authors)

  14. Investigation of evaluating the equipment for nuclear power plants with LCC

    International Nuclear Information System (INIS)

    Feng Chunping

    2010-01-01

    There are three stages for the equipment in nuclear power plants, they are production stage, service stage and remnant stage. The costs is relate to every stage in the whole life of the equipment. It also affects the economic performance of the nuclear power plant. Presently, when we evaluate the equipment during bidder program we mainly face to the production cost and ignore the service cost. So it can not evaluate the costs of the equipment in the whole life. All the costs shall be considered during the evaluation of the equipment to guarantee we can purchase the equipment completely and economically. (author)

  15. Evaluation of fire hazard analyses for nuclear power plants. A publication within the NUSS programme

    International Nuclear Information System (INIS)

    1995-01-01

    The present publication has been developed with the help of experts from regulatory, operating and engineering organizations, all with practical experience in the field of fire safety of nuclear power plants. The publication supplements the broad concepts of Safety Series No. 50-SG-D2 (Rev.1), Fire Protection in Nuclear Power Plants, by providing a detailed list of the issues, and some of the limitations, to be considered when evaluating the adequacy and effectiveness of the fire hazard analysis of a nuclear power plant. The publication is intended for assessors of fire hazard analyses, including regulators, independent assessors or plant assessors, and gives a broad description of the methodology to be used by operators in preparing a fire hazard analysis for their own plant. 1 fig

  16. Comparison of methods applicable to evaluation of nuclear power plant technical specifications

    International Nuclear Information System (INIS)

    Cho, N.Z.; Bozoki, G.E.; Youngblood, R.W.

    1986-01-01

    This study compares three probabilistic methods based on the static fault tree analysis, time-dependent unavailability analysis, and Markov analysis, which can be used to evaluate technical specifications in nuclear power plants. They are tested on a sample problem which was devised to closely represent the important and essential characteristics that should be addressed in determination and evaluation of the technical specifications

  17. A framework of risk-informed seismic safety evaluation of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Kondo, S.; Sakagami, M.; Hirano, M.; Shiba, M.

    2001-01-01

    A framework of risk-informed seismic design and safety evaluation of nuclear power plants is under consideration in Japan so as to utilize the progress in the seismic probabilistic safety assessment methodology. Issues resolved to introduce this framework are discussed after the concept, evaluation process and characteristics of the framework are described. (author)

  18. Analysis of difficulties accounting and evaluating nuclear material of PWR fuel plant

    International Nuclear Information System (INIS)

    Zhang Min; Jue Ji; Liu Tianshu

    2013-01-01

    Background: Nuclear materials accountancy must be developed for nuclear facilities, which is required by regulatory in China. Currently, there are some unresolved problems for nuclear materials accountancy of bulk nuclear facilities. Purpose: The retention values and measurement errors are analyzed in nuclear materials accountancy of Power Water Reactor (PWR) fuel plant to meet the regulatory requirements. Methods: On the basis of nuclear material accounting and evaluation data of PWR fuel plant, a deep analysis research including ratio among random error variance, long-term systematic error variance, short-term systematic error variance and total error involving Material Unaccounted For (MUF) evaluation is developed by the retention value measure in equipment and pipeline. Results: In the equipment pipeline, the holdup estimation error and its total proportion are not more than 5% and 1.5%, respectively. And the holdup estimation can be regraded as a constant in the PWR nuclear material accountancy. Random error variance, long-term systematic error variance, short-term systematic error variance of overall measurement, and analytical and sampling methods are also obtained. A valuable reference is provided for nuclear material accountancy. Conclusion: In nuclear material accountancy, the retention value can be considered as a constant. The long-term systematic error is a main factor in all errors, especially in overall measurement error and sampling error: The long-term systematic errors of overall measurement and sampling are considered important in the PWR nuclear material accountancy. The proposals and measures are applied to the nuclear materials accountancy of PWR fuel plant, and the capacity of nuclear materials accountancy is improved. (authors)

  19. Study on comprehensive evaluation model for nuclear power plant control room layout

    International Nuclear Information System (INIS)

    Zhu Yiming; Liu Yuan; Fan Huixian

    2010-01-01

    A comprehensive evaluation model for layout of the main control room of nuclear power plants was proposed. Firstly the design scope and principle for the layout of the main control room were defined based on the standards, and then the index system for the comprehensive evaluation was established. Finally, comprehensive evaluation was carried out for the layout design by applying the fuzzy comprehensive evaluation method in the index system. (authors)

  20. On nuclear power plant uprating

    International Nuclear Information System (INIS)

    Ho, S. Allen; Bailey, James V.; Maginnis, Stephen T.

    2004-01-01

    Power uprating for commercial nuclear power plants has become increasingly attractive because of pragmatic reasons. It provides quick return on investment and competitive financial benefits, while involving low risks regarding plant safety and public objection. This paper briefly discussed nuclear plant uprating guidelines, scope for design basis analysis and engineering evaluation, and presented the Salem nuclear power plant uprating study for illustration purposes. A cost and benefit evaluation of the Salem power uprating was also included. (author)

  1. Introduction of Integrity Evaluation Criteria Developing during and after fire for Nuclear Power Plant Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Woo [KEPCo, Gimcheon (Korea, Republic of)

    2016-05-15

    The first project for nuclear power plant built in Korea to taking account the engineering based approach was started on October 2015 including the whole process such as fire hazard analysis, standard fire modeling, cable tray fire with multi spurious operation, structural fire integrity evaluation, and large area fire induced air craft crash. This paper covers the brief developing scheme and roadmap focusing on structural fire evaluation criteria. The meaningful first step for developing the structural fire integrity in nuclear power plant building is started with the series of fire related sub sections mentioned in earlier section. The recognition and sufficient effort of fire research leads to set up the safe and reliable design of nuclear power plant.

  2. Development of resilience evaluation method for nuclear power plants. Part 3. Study on evaluation method and applicability of resilience index

    International Nuclear Information System (INIS)

    Suzuki, Masaaki; Demachi, Kazuyuki; Murakami, Kenta

    2016-01-01

    We have developed a new index, called the resilience index, that evaluates the dynamic stability of the system safety of a nuclear power plant during a severe accident by considering the ability to recover system safety functions that have become lost in the situation. In this paper, a detailed evaluation procedure for the resilience index is described. The system safety of a PWR plant during a severe accident is then assessed according to the resilience index in order to discuss the applicability of the index. We find that the resilience index successfully represents management capability and, therefore, the resilience capability of a nuclear power plant. (author)

  3. Digitized operator evaluation system for main control room of nuclear power plant

    International Nuclear Information System (INIS)

    Chen Yu; Yan Shengyuan; Chen Wenlong

    2014-01-01

    In order to evaluate the human-machine system matching relation of main control room in nuclear power plant accurately and efficiently, the expression and parameters of operator human body model were analyzed, and the evaluation required function of digital operator was determined. Based on the secondary development technology, the digital operator evaluation body model was developed. It could choose generation, gender, operation posture, single/eyes horizon, and left/right hand up to the domain according to the needs of specific evaluation, it was used to evaluate whether display information can be visible and equipment can be touch, and it also has key evaluation functions such as workspace and character visibility at the same time. The examples show that this method can complete the evaluation work of human-machine matching relation for main control room of nuclear power plant accurately, efficiently and quickly, and achieve the most optimal human-machine coordination relationship. (authors)

  4. IAEA effort on the evaluation and management of safety aspects of nuclear power plant ageing

    International Nuclear Information System (INIS)

    Pachner, J.; Yaremy, E.M.

    1991-01-01

    The questions attached to nuclear power plant (NPP) ageing and the need for monitoring and assessment of plant condition will grow in importance as more NPPs approach and pass the end of their nominal design lives. Ageing in nuclear plants must be effectively managed to ensure plant safety during their entire service life. This paper provides an overview of the IAEA programme and its results on the evaluation and management of safety aspects of NPP ageing. Under this programme, three generic guidance documents have been prepared on: data collection and record keeping; ageing management methodology; and the use of probabilistic safety assessment in plant life extension considerations. A two pronged strategy involving both technical and regulatory aspects has been adopted for the current and future work. (author)

  5. Development of web-based integrity evaluation system for primary components in a nuclear power plant

    International Nuclear Information System (INIS)

    Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J.; Choi, S.N.; Jang, K.S.; Hong, S.Y.

    2004-01-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

  6. Installation and evaluation of a nuclear power plant operator advisor based on artificial intelligence technology

    International Nuclear Information System (INIS)

    Hajek, B.K.; Miller, D.W.

    1989-01-01

    This report discusses the following topics on a Nuclear Power Plant operator advisor based on artificial Intelligence Technology; Workstation conversion; Software Conversion; V ampersand V Program Development Development; Simulator Interface Development; Knowledge Base Expansion; Dynamic Testing; Database Conversion; Installation at the Perry Simulator; Evaluation of Operator Interaction; Design of Man-Machine Interface; and Design of Maintenance Facility

  7. Projecting labor demand and worker immigration at nuclear power plant construction sites: an evaluation of methodology

    International Nuclear Information System (INIS)

    Herzog, H.W. Jr; Schlottmann, A.M.; Schriver, W.R.

    1981-12-01

    The study evaluates methodology employed for the projection of labor demand at, and worker migration to, nuclear power plant construction sites. In addition, suggestions are offered as to how this projection methodology might be improved. The study focuses on projection methodologies which forecast either construction worker migration or labor requirements of alternative types of construction activity. Suggested methodological improvements relate both to institutional factors within the nuclear power plant construction industry, and to a better use of craft-specific data on construction worker demand/supply. In addition, the timeliness and availability of the regional occupational data required to support, or implement these suggestions are examined

  8. Higher operational safety of nuclear power plants by evaluating the behaviour of operating personnel

    International Nuclear Information System (INIS)

    Mertins, M.; Glasner, P.

    1990-01-01

    In the GDR power reactors have been operated since 1966. Since that time operational experiences of 73 cumulative reactor years have been collected. The behaviour of operating personnel is an essential factor to guarantee the safety of operation of the nuclear power plant. Therefore a continuous analysis of the behaviour of operating personnel has been introduced at the GDR nuclear power plants. In the paper the overall system of the selection, preparation and control of the behaviour of nuclear power plant operating personnel is presented. The methods concerned are based on recording all errors of operating personnel and on analyzing them in order to find out the reasons. The aim of the analysis of reasons is to reduce the number of errors. By a feedback of experiences the nuclear safety of the nuclear power plant can be increased. All data necessary for the evaluation of errors are recorded and evaluated by a computer program. This method is explained thoroughly in the paper. Selected results of error analysis are presented. It is explained how the activities of the personnel are made safer by means of this analysis. Comparisons with other methods are made. (author). 3 refs, 4 figs

  9. Geographical evaluation of the impact of nuclear power plants on settlement structures

    International Nuclear Information System (INIS)

    Divinsky, B.

    1992-01-01

    The effects of nuclear power plants are classed with respect to their character (one-sided or many-sided), order (primary or secondary), quality (positive or adverse), duration (temporary or permanent), and space (microregional or macroregional). The following topics must be included in the methodology of evaluation of the impacts of a nuclear power plant on the region: characteristics of the present settlement network, relationships within the settlement system, spatial transformation of settlements, development of urbanization, population density, town and village sizes, functional types of settlements, migration, age and social structure of the population, economic activity, town and village facilities, technical infrastructure, transport and traffic, psycho-social impacts of the occurrence of the nuclear power plant, microecology (microenvironment). (M.D.). 5 refs

  10. Developments of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2003-03-15

    The objective of this research is to develop an efficient evaluation technology and to investigate applicability of newly-developed technology, such as internet-based cyber platform, to operating power plants. Development of efficient evaluation systems for Nuclear Power Plant components, based on structural integrity assessment techniques, are increasingly demanded for safe operation with the increasing operating period of Nuclear Power Plants. The following five topics are covered in this project: development of assessment method for wall-thinned nuclear piping based on pipe test; development of structural integrity program for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for mam components of NPP; development of internet-based cyber platform and integrity program for primary components of NPP; effect of aging on strength of dissimilar welds.

  11. The regulatory evaluation of radiation protection training programmes at Canadian nuclear power plants

    International Nuclear Information System (INIS)

    Legare, M.; Tennant, D.

    1996-01-01

    The responsibility for providing the necessary assurance that the use of nuclear energy in Canada does not pose undue risk to health, safety, security and the environment is vested with the Atomic Energy Control Board (AECB). This responsibility has led the Operator Certification Division of the AECB to develop methods to obtain assurance that nuclear power plant operations personnel are well trained and adequately competent to perform their duties. The features of the AECB approach to evaluation of training programmes based on a systematic approach to training is described. An overview of the Canadian nuclear power plants' radiation protection qualification levels is given. The developing evaluation process is contributing to the improvement of licensee radiation protection training programmes. This is making possible the transfer of part of the responsibility for licensed personnel radiation protection qualification assessment to the licensees, thus enabling a reduction in the operator certification division formal qualification activities. (author)

  12. A Study of KHNP Nuclear Power Plant Technology Level Evaluation

    International Nuclear Information System (INIS)

    Yang, Seung Han; Lee, Sung Jin; Kim, Yo Han

    2016-01-01

    KHNP's 2030 mid and long term plan goal in technology field is securing global No. 1 NPP technology level. Quantifying technology level for this purpose, technology level at present should be surveyed. Technology level of South Korea has been surveyed by KISTEP (Korea Institute of S and T Evaluation and Planning) every two year but the technology level of KHNP has not been surveyed by any organization including KHNP itself. Also the size of technology surveyed by KISTEP was too broad to quantifying technology level of KHNP. In this paper, technology level of KHNP and South Korea are presented. In this study, NPP related technologies were divided into Level I and Level II technologies and conducted a survey for each Level II technologies using Delphi questionnaire survey that is widely used in technology level evaluation. The results of technology level and gap will be used from strategic point of view and also as a reference data for technology improvement planning

  13. Development of human factors evaluation techniques for nuclear power plants

    International Nuclear Information System (INIS)

    Oh, I.S.; Lee, Y.H.; Lee, J.W.; Sim, B.S.

    1999-01-01

    This paper describes development of an operator task simulation analyzer and human factors evaluation techniques performed recently at Korea Atomic Energy Research Institute. The first is the SACOM (Simulation Analyzer with a Cognitive Operator Model) for the assessment of task performance by simulating control room operation. The latter has two objectives: to establish a human factors experiment facility, the Integrated Test Facility (ITF), and to establish techniques for human factors experiments. (author)

  14. Study on the safety evaluation method development for D and D of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sin, S. W.; Kang, G. D.; Kim, H. S.; Son, J. G.; Choi, Y. J.; Lee, K. J.; Koh, E. O.; Kim, K. D.; Ha, J. H. [Korea Hydro and Nuclear Power Co. Ltd., Seoul (Korea, Republic of)

    2004-02-15

    The final goal of decommissioning of Nuclear Power Plant is to remove or to isolate the radioactivity existing at the nuclear facilities which are confronted with termination of operation. The regulatory guidance and technologies of decommissioning have been developed strategically in some advanced countries. They have already stepped into the application stage from the research and development stage. The contents and the scope of this study is as follows, preparation of standard evaluation plan and establishment of standard safety evaluation system for D and D.

  15. Evaluation of mental workload on digital maintenance systems in nuclear power plants

    International Nuclear Information System (INIS)

    Hwang, S. L.; Huang, F. H.; Lin, J. C.; Liang, G. F.; Yenn, T. C.; Hsu, C. C.

    2006-01-01

    The purpose of this study is to evaluate operators' mental workload dealing with digital maintenance systems in Nuclear Power Plants. First of all, according to the factors affected the mental workload, a questionnaire was designed to evaluate the mental workload of maintenance operators at the second Nuclear Power (NPP) in Taiwan. Then, sixteen maintenance engineers of the Second NPP participated in the questionnaire survey. The results indicated that the mental workload was lower in digital systems than that in analog systems. Finally, a mental workload model based on Neural Network technique was developed to predict the workload of maintenance operators in digital maintenance systems. (authors)

  16. Handbook for the planning, co-ordination and evaluation of emergency exercises in nuclear power plants

    International Nuclear Information System (INIS)

    Schmidtborn, I.; Bath, N.

    1999-01-01

    The efficiency of the on-site emergency organization in German nuclear power plants is tested regularly through emergency exercises. To achieve federal harmonization on a high level of quality a handbook for the planning, co-ordination and evaluation of such exercises has been developed in the frame of the regulatory investigation programme. In this handbook requirements are set out for emergency training. Key elements are a modular structure, rules to be observed and guidance for post-exercise evaluation. (orig.) [de

  17. Economic evaluation of heat extraction from nuclear power plants - a criterion for deciding their building order

    International Nuclear Information System (INIS)

    Navratil, J.

    1987-01-01

    Heat extraction from nuclear power plants is an important element in the current concept of supplying the population and industries with heat. Economic evaluation of the extraction is one of the factors of the total economic assessment of potential sites for nuclear power plant construction which can contribute to decision making on the priorities of construction. The methodological approach to the assessment of economic contribution of heat extraction from 2x1000 MW nuclear power plant is exemplified using three such sites on the Czechoslovak territory, viz., Opatovice (eastern Bohemia), Blahutovice (northern Moravia), and Kecerovce (eastern Slovakia). The so-called annual converted cost was used as a suitable quantity completely reflecting all significant economic effects of heat extraction. It is shown that the fuel component of the power plant costs is the decisive factor for the amount of the annual converted cost in respect to heat supply and thus also the economic priority of the construction sites of nuclear power plants. (Z.M.). 3 tabs., 3 refs

  18. Study on integrity evaluation of structures associated with nuclear power plants

    International Nuclear Information System (INIS)

    2013-01-01

    The 3.11 Tohoku District -off the Pacific Ocean Earthquake and tsunami made us observations of tsunami height in large exceedance of the design, and besides it gave the most damages to several nuclear power plants facing the Pacific Ocean at source area of the earthquake. Particularly, at Fukushima-Daiichi Nuclear Power Plant, the great tsunami caused the simultaneous failure on several plant's equipment and components, which escalated into the core damage. Considering these background, the objective of this research is to enhance fundamental technology relative to integrity evaluation of SSC's (System, Structure, components) targeting external events such as earthquakes and tsunamis. Specifically, it is performed to develop structure evaluation methods against tsunami, to develop seismic isolation system, and to enhance non-liner analysis methods for building and so on. In viewpoint of the other external events except earthquake and tsunami, it was performed to develop impact analysis methods on building and outdoor structure against swept things caused by tornadoes. After that on the basis of these developments, it is performed to draw up guidelines such as the base isolation structure review guide, and the structure design and risk evaluation guide against tsunami, which are to be used in cross-check analysis targeting integrity evaluation of nuclear power plant's structures against external events such as earthquakes and tsunamis. (author)

  19. Evaluation of methods for seismic analysis of nuclear fuel reprocessing plants, part 1

    International Nuclear Information System (INIS)

    Tokarz, F.J.; Murray, R.C.; Arthur, D.F.; Feng, W.W.; Wight, L.H.; Zaslawsky, M.

    1975-01-01

    Currently, no guidelines exist for choosing methods of structural analysis to evaluate the seismic hazard of nuclear fuel reprocessing plants. This study examines available methods and their applicability to fuel reprocessing plant structures. The results of this study should provide a basis for establishing guidelines recommending methods of seismic analysis for evaluating future fuel reprocessing plants. The approach taken is: (1) to identify critical plant structures and place them in four categories (structures at or near grade; deeply embedded structures; fully buried structures; equipment/vessels/attachments/piping), (2) to select a representative structure in each of the first three categories and perform static and dynamic analysis on each, and (3) to evaluate and recommend method(s) of analysis for structures within each category. The Barnwell Nuclear Fuel Plant is selected as representative of future commercial reprocessing plants. The effect of site characteristics on the structural response is also examined. The response spectra method of analysis combined with the finite element model for each category is recommended. For structures founded near or at grade, the lumped mass model could also be used. If a time history response is required, a time-history analysis is necessary. (U.S.)

  20. Hazardous-material accidents near nuclear power plants: an evaluation of analyses and approaches

    International Nuclear Information System (INIS)

    Kot, C.A.; Lin, H.C.; van Erp, J.B.; Eichler, T.V.; Wiedermann, A.H.

    1983-10-01

    The state of knowledge concerning postulated accidents involving offsite hazardous materials in the vicinity of nuclear power plants is critically evaluated. This effort is part of a study to analyze the potential effects of offsite hazards upon the safety of nuclear power plants and to develop a technical basis for the assessment of siting approaches. The evaluation includes consideration of data bases and statistics of hazardous materials and accidents involving them, deterministic aspects of possible material dispersion and threat environments, the susceptibility and vulnerability of vital plant systems, and a critical review of past licensing experience and regulatory practice with respect to these hazards. While many of the data bases and analysis methods exist for an adequate estimate of threat and plant response, this knowledge is not fully used and no comprehensive guidance has been developed. Siting of nuclear power plants relative to offsite hazardous materials is a risk based procedure that considers both probabilities and consequences of events that make up accident scenarios. In this context it appears feasible to improve the procedures vis-a-vis the perception of safety, economy of effort, and efficiency of implementation. A scenario dependent conditional risk approach is outlined as a possible means of improving the siting procedures

  1. Evaluation of Haddam Neck (Connecticut Yankee) Nuclear Power Plant, environmental impact prediction, based on monitoring programs

    Energy Technology Data Exchange (ETDEWEB)

    Gore, K.L.; Thomas, J.M.; Kannberg, L.D.; Mahaffey, J.A.; Waton, D.G.

    1976-12-01

    A study was undertaken by the U.S. Nuclear Regulatory Commission (NRC) to evaluate the nonradiological environmental data obtained from three nuclear power plants operating for a period of one year or longer. The document presented reports the second of three nuclear power plants to be evaluated in detail by Battelle, Pacific Northwest Laboratories. Haddam Neck (Connecticut Yankee) Nuclear Power Plant nonradiological monitoring data were assessed to determine their effectiveness in the measurement of environmental impacts. Efforts were made to determine if: (1) monitoring programs, as designed, can detect environmental impacts, (2) appropriate statistical analyses were performed and if they were sensitive enough to detect impacts, (3) predicted impacts could be verified by monitoring programs, and (4) monitoring programs satisfied the requirements of the Environmental Technical Specifications. Both preoperational and operational monitoring data were examined to test the usefulness of baseline information in evaluating impacts. This included an examination of the methods used to measure ecological, chemical, and physical parameters, and an assessment of sampling periodicity and sensitivity where appropriate data sets were available. From this type of analysis, deficiencies in both preoperational and operational monitoring programs may be identified and provide a basis for suggested improvement.

  2. Guideline for the seismic technical evaluation of replacement items for nuclear power plants

    International Nuclear Information System (INIS)

    Harris, S.P.; Cushing, R.W.; Johnson, H.W.; Abeles, J.M.

    1993-02-01

    Seismic qualification for equipment originally installed in nuclear power plants was typically performed by the original equipment suppliers or manufactures (OES/OEM). Many of the OES/OEM no longer maintain quality assurance programs with adequate controls for supplying nuclear equipment. Utilities themselves must provide reasonable assurance in the continued seismic adequacy of such replacement items. This guideline provides practical, cost-effective techniques which can be used to provide reasonable assurance that replacement items will meet seismic performance requirements necessary to maintain the seismic design basis of commercial nuclear power plants. It also provides a method for determining when a seismic technical evaluation of replacement items (STERI) is required as part of the procurement process for spare and replacement items. Guidance on supplier program requirements necessary to maintain continued seismic adequacy and on documentation of maintaining required seismic adequacy is also included

  3. Nuclear plant scram reduction

    International Nuclear Information System (INIS)

    Wiegle, H.R.

    1986-01-01

    The Nuclear Utility Management and Human Resources Committee (NUMARC) is a confederation of all 55 utilities with nuclear plants either in operation or under construction. NUMARC was formed in April 1984 by senior nuclear executives with hundreds of man-years of plant experience to improve (plant) performance and resolve NRC concerns. NUMARC has adopted 10 commitments in the areas of management, training, staffing and performance. One of these commitments is to strive to reduce automatic trips to 3 per year per unit for calendar year 1985 for plants in commercial operation greater than 3 years (with greater than 25% capacity factor). This goal applies to any unplanned automatic protection system trips at any time when the reactor is critical. Each utility has committed to develop methods to thoroughly evaluate all unplanned automatic trips to identify the root causes and formulate plans to correct the root causes thus reducing future unplanned scrams. As part of this program, the Institute of Nuclear Power Operations (INPO) collects and evaluates information on automatic reactor trips. It publishes the results of these evaluations to aid the industry to identify root causes and corrective actions

  4. Method on the aging evaluation in nuclear power plant concrete structures

    International Nuclear Information System (INIS)

    Kitsutaka, Yoshinori; Tsukagoshi, Masayuki

    2014-01-01

    In this paper, method on the durability evaluation in nuclear power plant concrete structures was investigated. In view of the importance of evaluating the degree of deterioration of reinforced concrete structures, relationships should be formulated among the number of years elapsed, t, the amount of action of a deteriorative factor, F, the degree of material deterioration, D, and the performance of the structure, P. Evaluation by PDFt diagrams combining these relationships may be effective. A detailed procedure of durability evaluation for a reinforced concrete structure using PDFt concept is presented for the deterioration factors of thermal effect, irradiation, neutralization and penetration of salinity by referring to the recent papers

  5. Evaluation issues on real-time operating system in nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Y. M.; Jeong, C. H.; Koh, J. S.

    2006-01-01

    In the recent few years, using the hard real-time operating system (RTOS) of safety-critical applications has gained increased acceptance in the nuclear safety system. Failure of this software could cause catastrophic consequences for human life. The digital I and C systems of nuclear power plants also have used hard RTOSs which are executing a required mission completely within its deadline. Because the nuclear power plants have to maintain a very high level of safety, the hard RTOS software should be reliable and safe. The RTOS used in safety-critical I and C systems is the base software used for the purpose of satisfying the real-time constraints, So, careful evaluation of its safety and functionality is very important, So far, the nuclear power plants of Korea have adopted commercial off-the-shelf (COTS) RTOS software. But, these days the RTOS embedded in safety grade PLC has been developed by KNICS project controlled by Ministry of Commerce, Industry and Energy of Korea. Whether COTS RTOS or newly developed RTOS, it must be evaluated its safety and reliability. (authors)

  6. Geotechnical aspects of site evaluation and foundations for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    This publication is a revision of the former safety standards of IAEA Safety Series No. 50-SG-S8. The scope has been extended to cover not only foundations but also design questions related to geotechnical science and engineering, such as the bearing capacity of foundations, design of earth structures and design of buried structures Seismic aspects also play an important role in this field, and consequently the Safety Guide on Evaluation of Seismic Hazards for Nuclear Power Plants, Safety Standards Series No. NS-G-3.3, which discusses the determination of seismic input motion, is referenced on several occasions. The present Safety Guide provides an interpretation of the Safety Requirements on Site Evaluation for Nuclear Installations and guidance on how to implement them. It is intended for the use of safety assessors or regulators involved in the licensing process as well as the designers of nuclear power plants, and it provides them with guidance on the methods and procedures for analyses to support the assessment of the geotechnical aspects of the safety of nuclear power plants

  7. Geotechnical aspects of site evaluation and foundations for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2003-01-01

    This publication is a revision of the former safety standards of IAEA Safety Series No. 50-SG-S8. The scope has been extended to cover not only foundations but also design questions related to geotechnical science and engineering, such as the bearing capacity of foundations, design of earth structures and design of buried structures. Seismic aspects also play an important role in this field, and consequently the Safety Guide on Evaluation of Seismic Hazards for Nuclear Power Plants, Safety Standards Series No. NS-G-3.3, which discusses the determination of seismic input motion, is referenced on several occasions. The present Safety Guide provides an interpretation of the Safety Requirements on Site Evaluation for Nuclear Installations and guidance on how to implement them. It is intended for the use of safety assessors or regulators involved in the licensing process as well as the designers of nuclear power plants, and it provides them with guidance on the methods and procedures for analyses to support the assessment of the geotechnical aspects of the safety of nuclear power plants

  8. Geotechnical aspects of site evaluation and foundations for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This publication is a revision of the former safety standards of IAEA Safety Series No. 50-SG-S8. The scope has been extended to cover not only foundations but also design questions related to geotechnical science and engineering, such as the bearing capacity of foundations, design of earth structures and design of buried structures. Seismic aspects also play an important role in this field, and consequently the Safety Guide on Evaluation of Seismic Hazards for Nuclear Power Plants, Safety Standards Series No. NS-G-3.3, which discusses the determination of seismic input motion, is referenced on several occasions. The present Safety Guide provides an interpretation of the Safety Requirements on Site Evaluation for Nuclear Installations and guidance on how to implement them. It is intended for the use of safety assessors or regulators involved in the licensing process as well as the designers of nuclear power plants, and it provides them with guidance on the methods and procedures for analyses to support the assessment of the geotechnical aspects of the safety of nuclear power plants

  9. Summary of plant life management evaluation for Onagawa Nuclear Power Station Unit-1

    International Nuclear Information System (INIS)

    Nodate, Kazumi

    2014-01-01

    The Onagawa Nuclear Power Station Unit-1 (Onagawa NPS-1) began commercial operation on June 1, 1984, and has reached 30-year from starting of operation on June of 2014. To that end, we implemented the Plant Life Management (PLM) evaluation for Onagawa NPS-1 as our first experience. We decided on a Long-term Maintenance Management Policy from result of the evaluation, and then applied the Safety-Regulations change approval application on November 6, 2013 and its correcting application on April 16, 2014. Our application was approved on May 21, 2014 through investigation by the Nuclear Regulatory Agency. Also at implementation of the PLM evaluation, we considered effects of the Great East Japan Earthquake that occurred on March 11, 2011 against ageing phenomena. In this paper, we introduce summary of PLM evaluation for Onagawa NPS-1 and the evaluation that considered effects of the Great East Japan Earthquake. (author)

  10. An improved basis for evaluating continued service of Category I concrete structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Mori, Y.; Arndt, E.G.

    1992-01-01

    The Structural Aging (SAG) Program has the overall objective of preparing technical bases for regulatory criteria which will provide the NRC with potential structural safety issues and acceptance criteria for use in nuclear power plant evaluations for continued service. In meeting this objective three primary activities are underway: (1) development of a structural materials information center containing data and information on the variation of concrete and concrete-related material properties over time; (2) establishment of procedures to make quantitative evaluations of the presence, magnitude, and significance of environmental stressors or aging factors that can impact critical component performance, as well as techniques which can be used for repair of degraded concrete structures; and (3) formulation of a quantitative methodology for performing current condition assessments and making reliability-based life predictions of critical concrete structures in nuclear power plants. Accomplishments to date under each of these tasks are presented

  11. Evaluation of digital fault-tolerant architectures for nuclear power plant control systems

    International Nuclear Information System (INIS)

    Battle, R.E.

    1990-01-01

    This paper reports on four fault-tolerant architectures that were evaluated for their potential reliability in service as control systems of nuclear power plants. The reliability analyses showed that human- and software-related common cause failures and single points of failure in the output modules are dominant contributors to system unreliability. The four architectures are triple-modular-redundant, both synchronous and asynchronous, and also dual synchronous and asynchronous. The evaluation includes a review of design features, an analysis of the importance of coverage, and reliability analyses of fault-tolerant systems. Reliability analyses based on data from several industries that have fault-tolerant controllers were used to estimate the mean-time-between-failures of fault-tolerant controllers and to predict those failure modes that may be important in nuclear power plants

  12. Evaluation of the emergency warning system at the Fort St. Vrain nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Sorensen, J.H.

    1982-11-01

    The Fort St. Vrain power plant is the only high-temperature gas-cooled reactor (HTGR) in commercial operation in the United States. All commercial reactors, regardless of technology, must conform to Nuclear Regulatory Commission emergency planning regulations developed in light of Clarification of TMI Action Plan Requirements (NUREG-0737). This report analyzes the applicability of warning-related planning requirements to HTGRs and evaluates the strengths and weaknesses of warning procedures at Fort St. Vrain.

  13. Evaluation of Millstone Nuclear Power Plant, Environmental Impact prediction, based on monitoring programs

    Energy Technology Data Exchange (ETDEWEB)

    Gore, K.L.; Thomas, J.M.; Kannberg, L.D.; Watson, D.G.

    1977-02-01

    This report evaluates the nonradiological monitoring program at Millstone Nuclear Power Plant. Both operational as well as preoperational monitoring programs were analyzed to produce long-term (5 yr or longer) data sets, where possible. In order to determine the effectiveness of these monitoring programs, the appropriate data sets have to be analyzed by the appropriate statistical analysis. Thus, both open literature and current statistical analysis being developed at Pacific Northwest Laboratories (PNL) were employed in data analysis.

  14. Stress corrosion evaluation on stainless steel 304 pipes in Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Arganis J, C.R.

    1996-01-01

    Inside the frame of the project IAEA/MEX-41044 'Stress corrosion as a starting event of accidents in nuclear plants', and of the institutional project IA-252 under the same name, it was required from the Laguna Verde Nuclear Plant, material equivalent to the one employed in the piping of the primary recycling system. Laguna Verde Nuclear Plant granted two tracks of tubes, that could be used to substitute the ones that are in operation, as is the tube SA-358TP304 CL-QC with transversal welding, designated as ER-316-LQA. According to the report entitles 'Revision of the operational experience related to corrosion in the nuclear plants' it was found that the stress corrosion is the principal mechanism of corrosion present in the nuclear plants. Previous records indicate that sensitized stainless steels are resistant to stress corrosion in testings of constant loading in sea water (3.5% of chlorides approximately) to 80 Centigrade and to 80% of the limit of conveyance and that a solution of 22% of NaCl to 90 Centigrade, produces cracking due to stress corrosion in highly sensitized steels, in tests of speed of slow extension (SSRT), to a speed of 1x10 -6 s -1 . Daniels reports that there is a direct relation between the speed limit of detection of the SSRT test and the concentration of chlorides, for stainless steels tested to 100 Centigrade. The minimum detection speed of susceptibility to stress corrosion for solution to 20% of NaCl, is of 1x10 -7 s -1 . Taking into account these considerations, the employment of a solution with 22% of NaCl to 90 Centigrade to a speed of 1x10 -6 s -1 seems a good choice for the evaluation of stainless steel. (Author)

  15. Cyber Security Evaluation of the Wireless Communication for the Mobile Safeguard Systems in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, S.; Kim, Y.S.; Ye, S.H.

    2015-01-01

    This paper introduces cyber security evaluation results and a design of the wireless communication technology to apply to safeguard systems in nuclear power plants. While wireless communication technologies can generally make mobility and efficiency on plant operation, those have seldom been installed on the nuclear I&C systems due to the negative concern of unexpected outcomes that stem from electromagnetic interference and cyber attack. New design of advanced digital safeguard and I&C systems uses computer-based systems for the safeguard and safety functions. On the other hand, those are being exposed to various types of new and existing cyber threats, vulnerabilities and risks which significantly increase the likelihood that those could be compromised. In order to employ the wireless communication technology in safeguard function, licencees assess and manage the potential for adverse effects on safeguard and safety functions so as to provide high assurance that critical functions are properly protected cyber attack. It is expected that the safeguard function, specifically on the area of real-time monitoring, logging, can be enhanced by employing the mobile safeguard devices (: smart phone, laptop, smart pad, etc). In this paper, we deal with the cyber security evaluation, which consists of threat analysis, vulnerability test, establishment of security plan, and design solutions for the wireless communication on the basis of IEEE 802.11(Wi-Fi) protocol. Proposed evaluation and design solution could be a basis for the design of wireless communication and mobile safeguard systems in nuclear power plants. (author)

  16. Evaluation of safety issues on newly regulated nuclear power plant by tsunami-level 1 PRA

    International Nuclear Information System (INIS)

    Tsuji, Yutaro; Miwa, Shuichiro; Mori, Michitsugu

    2014-01-01

    The tsunami caused by the Great East Japan Earthquake triggered severe accidents involving the units 1 to 4 at the Fukushima Dai-ichi nuclear power station (NPS). In order to re-operate existing nuclear power plants it should be necessary to reduce the core damage frequency on risk by tsunami. In this work, effects of the off-site power supply installation on resuming operation of nuclear power plants were investigated by utilizing the Tsunami-Level 1 Probability Risk Assessment (PRA). Unit 2 of the Onagawa nuclear power station, which resembled units 2 and 3 of Fukushima Dai-ichi, was selected for PRA. First, event-tree was created for the units of the Onagawa nuclear power station with the safety systems such as Emergency Core Cooling System (ECCS), investigating the plant situation at the time of the earthquake and tsunami occurrences. It was assumed that the magnitude of the tsunami was equivalent to the Great East Japan Earthquake. The accident-analytical progression-time was 36 hours, determined from the core-damage occurrence of the unit 3 of Fukushima Dai-ichi nuclear power station. Failure probabilities were calculated by the fault tree, which was created from the elements listed in the event tree. For the calculation, failure rates reported by the NUCIA (NUClear Information Archives) were primarily utilized. Then, obtained failure probabilities were embedded to the event tree. Core damage probabilities were evaluated by calculating success and failure rates for each accidental progression and scenarios. Restoration of the failed equipment and machineries was not considered in the analysis. Installation of the power supply vehicles at the nuclear power plant site reduced the core damage probability from 2.58×10 -6 to 8.56×10 -7 . However, continued addition of the power supply vehicles could not lower the core damage probability further more. In the case of Unit 2 of Onagawa nuclear power station, there could be a limit to lower the core damage

  17. A personal computer code for seismic evaluations of nuclear power plant facilities

    International Nuclear Information System (INIS)

    Xu, J.; Graves, H.

    1990-01-01

    A wide range of computer programs and modeling approaches are often used to justify the safety of nuclear power plants. It is often difficult to assess the validity and accuracy of the results submitted by various utilities without developing comparable computer solutions. Taken this into consideration, CARES is designed as an integrated computational system which can perform rapid evaluations of structural behavior and examine capability of nuclear power plant facilities, thus CARES may be used by the NRC to determine the validity and accuracy of analysis methodologies employed for structural safety evaluations of nuclear power plants. CARES has been designed to: operate on a PC, have user friendly input/output interface, and have quick turnaround. The CARES program is structured in a modular format. Each module performs a specific type of analysis. The basic modules of the system are associated with capabilities for static, seismic and nonlinear analyses. This paper describes the various features which have been implemented into the Seismic Module of CARES version 1.0. In Section 2 a description of the Seismic Module is provided. The methodologies and computational procedures thus far implemented into the Seismic Module are described in Section 3. Finally, a complete demonstration of the computational capability of CARES in a typical soil-structure interaction analysis is given in Section 4 and conclusions are presented in Section 5. 5 refs., 4 figs

  18. Evaluation of barriers and resilience to improve organizational performance in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, P. F.; Martin del Campo, C., E-mail: pnelson_007@yahoo.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec 62550, Morelos (Mexico)

    2014-10-15

    In this study, several models are built from the information contained in the Condition Reports from the Corrective Action Program ar a nuclear power plant. Condition Reports are seen as indications of organizational stress levels with consequential Condition Reports being further evidence of organizational resilience being exceeded. Also contained in this paper is an examination of methods used to assess organizational resilience thresholds to preclude consequential events, based on examination of the number and severity of the Condition Reports. In combination with PSA risk insights, it is possible to identify risk significant procedures and their corresponding risk significant steps. Leading indicators can be used to identify the need for installing a barrier or defense to reduce human errors in a nuclear power plant. These indicators are developed from the Corrective Action Program data by detecting increases in events. Organizational barriers can then be identified to improve performance. The resulting identified barriers are evaluated to rank the value of each possible barrier, and determine the best barrier(s) to implement. The tool described in this paper is designed to provide a systematic approach to identify areas where improvements in organizational effectiveness best reduce the likelihood of consequential events. Due to the uncertainty of many of the factors that influence the performance of humans in nuclear power plant activities, we propose using Bayesian networks to identify sources of organizational errors leading to consequential events. This study, using actual nuclear power plant data, includes a method for data processing and highlights some potential uses of Bayesian networks for improving organizational effectiveness in the nuclear power industry. (Author)

  19. Evaluation of barriers and resilience to improve organizational performance in nuclear power plants

    International Nuclear Information System (INIS)

    Nelson, P. F.; Martin del Campo, C.

    2014-10-01

    In this study, several models are built from the information contained in the Condition Reports from the Corrective Action Program ar a nuclear power plant. Condition Reports are seen as indications of organizational stress levels with consequential Condition Reports being further evidence of organizational resilience being exceeded. Also contained in this paper is an examination of methods used to assess organizational resilience thresholds to preclude consequential events, based on examination of the number and severity of the Condition Reports. In combination with PSA risk insights, it is possible to identify risk significant procedures and their corresponding risk significant steps. Leading indicators can be used to identify the need for installing a barrier or defense to reduce human errors in a nuclear power plant. These indicators are developed from the Corrective Action Program data by detecting increases in events. Organizational barriers can then be identified to improve performance. The resulting identified barriers are evaluated to rank the value of each possible barrier, and determine the best barrier(s) to implement. The tool described in this paper is designed to provide a systematic approach to identify areas where improvements in organizational effectiveness best reduce the likelihood of consequential events. Due to the uncertainty of many of the factors that influence the performance of humans in nuclear power plant activities, we propose using Bayesian networks to identify sources of organizational errors leading to consequential events. This study, using actual nuclear power plant data, includes a method for data processing and highlights some potential uses of Bayesian networks for improving organizational effectiveness in the nuclear power industry. (Author)

  20. Development of a methodology for the evaluation of radiation protection performance and management in nuclear power plants

    International Nuclear Information System (INIS)

    Schieber, Caroline; Bataille, Celine; Cordier, Gerard; Delabre, Herve; Jeannin, Bernard

    2008-01-01

    This paper describes a specific methodology adopted by Electricite de France to perform the evaluation of radiation protection performance and management within its 19 nuclear power plants. The results obtained in 2007 are summed up. (author)

  1. Submarine nuclear power plants

    International Nuclear Information System (INIS)

    Horton, C.C.

    1984-01-01

    The submarine nuclear power plant has revolutionised the strategy and tactics of under-sea warfare. Present day submarine nuclear power plants are discussed, as well as future developments. The endurance, speed, noise and diving depth of nuclear submarines are also outlined. (U.K.)

  2. Consideration of Evaluation of Communication using Work Domain Analysis (WDA) in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jang, In Seok; Seong, Poong Hyun; Park, Jin Kyun

    2009-01-01

    The nature of work has changed, this corresponding to a trend toward to computerization. In this phase, the role of people began to evolve from one of manual laborer, requiring primarily perception-motor skill, to intellectual worker, thereby requiring more conceptual knowledge and cognitive skills which means system such as nuclear power plant are getting more complicated and complex. Thus, the evolution of work has led to a greater demand for communication, collaboration, problem solving thereby increasing the discretion, and therefore the variability, in worker action. Because of these reasons, traditional approaches, normative method and descriptive method, have not been proper anymore. Naikar points out that by focusing on constraints, rather than on particular ways of working, it is possible to support workers in adapting their behavior online and in real time in a variety of situation, including unanticipated events. For these complex domain such as communication in nuclear power plant control room, an approach is required that models the conditions framing formative behavior, allowing the examination of emergent, unanticipated, unpredicted actions. In this study, it could be helpful to introduce the method that is proper to apply in complex and unanticipated like nuclear power plants. Thus, Abstraction Decomposition Space (ADS) which is the tool of Work Domain Analysis(WDA) is presented as an approach that is particularly amenable for this domain. The aim is to address ADS as a beginning of modeling the structure of what need to be analyzed can be used to support the analysis of communication in nuclear power plants. If the model that is made by ADS is correct, quantitative evaluation of communication could be done

  3. Regulatory evaluation of nuclear power plant on-the-job training programmes

    International Nuclear Information System (INIS)

    Wilkinson, J.

    1996-01-01

    It is now well recognized internationally in the nuclear industry that a Systematic Approach to Training (SAT), as described in IAEA-TECDOC-525, represents the currently most effective method of providing effective, efficient training to Nuclear Power Plant (NPP) personnel. For this reason the Canadian regulator, the Atomic Energy Control Board (AECB), has made SAT a requirement for the training of NPP personnel whose job activities could affect plant or public safety. In this respect the AECB recognizes that the SAT process will dictate on-the-job-training (OJT) and the prudent use of Job Performance Measures (JPMs) as the most effective method of providing parts of the required training to some work groups. The AECB has adopted a process of training program evaluation for NPP operations personnel which comprises methods to evaluate all types of training including OJT. This paper describes some variations which have been identified in the conduct of OJT at Canadian utilities. It then presents the three step process established by the AECB to effectively evaluate training programs expertise is required. The concept of utility self-evaluations is introduced. Finally, the importance of consistency on the part of the regulator in following a systematic approach to evaluation through the application of a viable standard is addressed. (author). 2 figs

  4. Preliminary evaluation of the profitability indexes of the Laguna Verde nuclear power plant

    International Nuclear Information System (INIS)

    Villanueva M, C.

    2010-10-01

    The Laguna Verde nuclear power plant of the Federal Commission of Electricity has an installed capacity of 1,350 MW and unit 1 started commercial operation in 1990 and unit 2 in 1995. This paper is a synthesis of the results of a preliminary evaluation of the expected profitability indexes of the power plant during an economic lifetime of 40 years. The following data was used as input to the evaluation model prescribed by the Finance and Public Credit Secretary for public investment projects. Unit investment cost: 3,500 US D/k W; Fixed operation and maintenance cost: 54. 45 US D/year-k W; Variable operation and maintenance cost: 0. 38 US D/M Wh; Nuclear fuel cycle cost: 10. 28 US D/M Wh; Lifetime capacity factor: 85%; Discount rate: 12.0% per year; Sale price of electricity to the interconnected electric system: 80. 75 US D/M Wh. The output of the evaluation model is the following: Cost of electricity generated: 60. 2 1 US D/M Wh; fixed cost 49. 55 US D/M Wh; variable cost 10. 66 US D/M Wh; Internal rate of return (Irr): 18.0%; Benefit to cost quotient (B/C): 1.341. A very systematic sensitivity analysis was done, that shows that the cost is very sensitive to the capacity factor and to the investment cost, but is very insensitive to the fixed operation and maintenance cost and to the nuclear fuel cost. Finally, a comparison was made to the evaluation of the profitability indexes of a natural gas fired combined cycle power plant. (Author)

  5. An estimate and evaluation of design error effects on nuclear power plant design adequacy

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1984-01-01

    An area of considerable concern in evaluating Design Control Quality Assurance procedures applied to design and analysis of nuclear power plant is the level of design error expected or encountered. There is very little published data 1 on the level of error typically found in nuclear power plant design calculations and even less on the impact such errors would be expected to have on overall design adequacy of the plant. This paper is concerned with design error associated with civil and mechanical structural design and analysis found in calculations which form part of the Design or Stress reports. These reports are meant to document the design basis and adequacy of the plant. The estimates contained in this paper are based on the personal experiences of the author. In Table 1 is a partial listing of the design docummentation review performed by the author on which the observations contained in this paper are based. In the preparation of any design calculations, it is a utopian dream to presume such calculations can be made error free. The intent of this paper is to define error levels which might be expected in a competent engineering organizations employing currently technically qualified engineers and accepted methods of Design Control. In addition, the effects of these errors on the probability of failure to meet applicable design code requirements also are estimated

  6. Performance-based evaluation of graphic displays for nuclear-power-plant control rooms

    International Nuclear Information System (INIS)

    Petersen, R.J.; Banks, W.W.; Gertman, D.I.

    1982-01-01

    This paper reports several methodologies for evaluating the perceptual and perceptual/decision making aspects of displays used in the control rooms of nuclear power plants. This NRC funded study focuses upon the Safety Parameter Display System (SPDS) and relates the utility of the display to objective performance and preference measures obtained in experimental conditions. The first condition is a traditional laboratory setting where classical experimental methodologies can be employed. The second condition is an interactive control room simulation where the operator's performance is assessed while he/she operates the simulator. The third condition is a rating scale designed to assess operator preferences and opinions regarding a variety of display formats. The goal of this study is the development of a cost-efficient display evaluation methodology which correlates highly with the operator's ability to control a plant

  7. A Study on the Evaluation Technology of Welds Integrity in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Chang, Hyun Young; Kim, Jong Sung; Jin, Tae Eun

    2007-01-01

    The final goal of this study is to develop the core technologies applicable to the design, operation and maintenance of welds in nuclear structures. This study includes predicting microstructure changes and residual stress for welded parts of nuclear power plant components. Furthermore, researches are performed on evaluating fatigue, corrosion, and hydrogen induced cracking and finally constructs systematically integrated evaluation system for structural integrity of nuclear welded structures. In this study, metallurgical and mechanical approaches have been effectively coordinated considering real welding phenomena in the fields of welds properties such as microstructure, composition and residual stress, and in the fields of damage evaluations such as fatigue, corrosion, fatigue crack propagation, and stress corrosion cracking. Evaluation techniques tried in this study can be much economical and effective in that it uses theoretical/semi-empirical but includes many additional parameters that can be introduced in real phenomena such as phase transformation, strength mismatch and residual stress. It is clear that residual stress makes great contribution to fatigue and stress corrosion cracking. Therefore the mitigation techniques have been approached by reducing the residual stress of selected parts resulting in successful conclusions

  8. Development and evaluation of tracking method for augmented reality system for nuclear power plant maintenance support

    International Nuclear Information System (INIS)

    Ishii, Hirotake; Bian, Zhiqiang; Sekiyama, Tomoki; Shimoda, Hiroshi; Yoshikawa, Hidekazu; Izumi, Masanori; Kanehira, Yoshiki; Morishita, Yoshitsugu

    2006-01-01

    This study aims at developing an augmented reality system to support maintenance work of nuclear power plants. An accurate and wide-range tracking method is required as a key technology in order to realize the system. In this study, a new tracking method using multi-camera and gyro sensor has been developed in order to enlarge the area where the tracking is available with limited number of markers. Experimental evaluation result shows that the area where the developed method can cover is about 3 times larger than the method using single camera. (author)

  9. Development and evaluation of tracking method for augmented reality system for nuclear power plant maintenance support

    International Nuclear Information System (INIS)

    Ishii, Hirotake; Bian, Zhiqiang; Sekiyama, Tomoki; Shimoda, Hiroshi; Yoshikawa, Hidekazu; Izumi, Masanori; Kanehira, Yoshiki; Morishita, Yoshitsugu

    2007-01-01

    This study aims at developing an augmented reality system to support maintenance work of nuclear power plants. An accurate and wide-range tracking method is required as a key technology in order to realize the system. In this study, a new tracking method using multi-camera and gyro sensor has been developed in order to enlarge the area where the tracking is available with limited number of markers. Experimental evaluation result shows that the area where the developed method can cover is about 3 times larger than the method using only single camera. (author)

  10. Images of nuclear power plants

    International Nuclear Information System (INIS)

    Hashiguchi, Katsuhisa; Misumi, Jyuji; Yamada, Akira; Sakurai, Yukihiro; Seki, Fumiyasu; Shinohara, Hirofumi; Misumi, Emiko; Kinjou, Akira; Kubo, Tomonori.

    1995-01-01

    This study was conducted to check and see, using Hayashi's quantification method III, whether or not the respondents differed in their images of a nuclear power plant, depending on their demographic variables particularly occupations. In our simple tabulation, we compared subject groups of nuclear power plant employees with general citizens, nurses and students in terms of their images of a nuclear power plant. The results were that while the nuclear power plant employees were high in their evaluations of facts about a nuclear power plant and in their positive images of a nuclear power plant, general citizens, nurses and students were overwhelmingly high in their negative images of a nuclear power plant. In our analysis on category score by means of the quantification method III, the first correlation axis was the dimension of 'safety'-'danger' and the second correlation axis was the dimension of 'subjectivity'-'objectivity', and that the first quadrant was the area of 'safety-subjectivity', the second quadrant was the area of 'danger-subjectivity', the third quadrant as the area of 'danger-objectivity', and the forth quadrant was the area of 'safety-objectivity'. In our analysis of sample score, 16 occupation groups was compared. As a result, it was found that the 16 occupation groups' images of a nuclear power plant were, in the order of favorableness, (1) section chiefs in charge, maintenance subsection chiefs, maintenance foremen, (2) field leaders from subcontractors, (3) maintenance section members, operation section members, (4) employees of those subcontractors, (5) general citizens, nurses and students. On the 'safety-danger' dimension, nuclear power plant workers on the one hand and general citizens, nurses and students on the other were clearly divided in terms of their images of a nuclear power plant. Nuclear power plant workers were concentrated in the area of 'safety' and general citizens, nurses and students in the area of 'danger'. (J.P.N.)

  11. Nuclear power plant outages

    International Nuclear Information System (INIS)

    1998-01-01

    The Finnish Radiation and Nuclear Safety Authority (STUK) controls nuclear power plant safety in Finland. In addition to controlling the design, construction and operation of nuclear power plants, STUK also controls refuelling and repair outages at the plants. According to section 9 of the Nuclear Energy Act (990/87), it shall be the licence-holder's obligation to ensure the safety of the use of nuclear energy. Requirements applicable to the licence-holder as regards the assurance of outage safety are presented in this guide. STUK's regulatory control activities pertaining to outages are also described

  12. System model for evaluation of an emergency response plan for a nuclear power plant based on an assessment of nuclear emergency exercises

    International Nuclear Information System (INIS)

    Silva, Marcos Vinicius C.; Medeiros, Jose A.C.C.

    2011-01-01

    Nuclear power plants are designed and built with systems dedicated to provide a high degree of protection to its workers, the population living in their neighborhoods and the environment. Among the requirements for ensuring safety there are the existence of the nuclear emergency plan. Due to the relationship between the actions contemplated in the emergency plan and the nuclear emergency exercise, it becomes possible to assess the quality of the nuclear emergency plan, by means of emergency exercise evaluation, The techniques used in this work aim at improving the evaluation method of a nuclear emergency exercise through the use of performance indicators in the evaluation of the structures, actions and procedures involved. The proposed model enables comparisons between different moments of an emergency plan directed to a nuclear power plant as well as comparisons between plans dedicated to different facilities. (author)

  13. System model for evaluation of an emergency response plan for a nuclear power plant based on an assessment of nuclear emergency exercises

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Marcos Vinicius C.; Medeiros, Jose A.C.C. [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    Nuclear power plants are designed and built with systems dedicated to provide a high degree of protection to its workers, the population living in their neighborhoods and the environment. Among the requirements for ensuring safety there are the existence of the nuclear emergency plan. Due to the relationship between the actions contemplated in the emergency plan and the nuclear emergency exercise, it becomes possible to assess the quality of the nuclear emergency plan, by means of emergency exercise evaluation, The techniques used in this work aim at improving the evaluation method of a nuclear emergency exercise through the use of performance indicators in the evaluation of the structures, actions and procedures involved. The proposed model enables comparisons between different moments of an emergency plan directed to a nuclear power plant as well as comparisons between plans dedicated to different facilities. (author)

  14. Evaluation and Numerical Simulation of Tsunami for Coastal Nuclear Power Plants of India

    International Nuclear Information System (INIS)

    Sharma, Pavan K.; Singh, R.K.; Ghosh, A.K.; Kushwaha, H.S.

    2006-01-01

    Recent tsunami generated on December 26, 2004 due to Sumatra earthquake of magnitude 9.3 resulted in inundation at the various coastal sites of India. The site selection and design of Indian nuclear power plants demand the evaluation of run up and the structural barriers for the coastal plants: Besides it is also desirable to evaluate the early warning system for tsunami-genic earthquakes. The tsunamis originate from submarine faults, underwater volcanic activities, sub-aerial landslides impinging on the sea and submarine landslides. In case of a submarine earthquake-induced tsunami the wave is generated in the fluid domain due to displacement of the seabed. There are three phases of tsunami: generation, propagation, and run-up. Reactor Safety Division (RSD) of Bhabha Atomic Research Centre (BARC), Trombay has initiated computational simulation for all the three phases of tsunami source generation, its propagation and finally run up evaluation for the protection of public life, property and various industrial infrastructures located on the coastal regions of India. These studies could be effectively utilized for design and implementation of early warning system for coastal region of the country apart from catering to the needs of Indian nuclear installations. This paper presents some results of tsunami waves based on different analytical/numerical approaches with shallow water wave theory. (authors)

  15. Safety indicators as a tool for operational safety evaluation of nuclear power plants

    International Nuclear Information System (INIS)

    Araujo, Jefferson Borges; Melo, Paulo Fernando Ferreira Frutuoso e; Schirru, Roberto

    2009-01-01

    Performance indicators have found a wide use in the conventional and nuclear industries. For the conventional industry, the goal is to optimize production, reducing loss of time with accidents, human error and equipment downtimes. In the nuclear industry, nuclear safety is an additional goal. This paper presents a general methodology to the establishment, selection and use of safety indicators for a two loop PWR plant, as Angra 1. The use of performance indicators is not new. The NRC has its own methodology and the IAEA presents methodology suggestions, but there is no detailed documentation about indicators selection, criteria and bases used. Additionally, only the NRC methodology performs a limited integrated evaluation. The study performed identifies areas considered critical for the plant operational safety. For each of these areas, strategic sub-areas are defined. For each strategic sub-area, specific safety indicators are defined. These proposed Safety Indicators are based on the contribution to risk considering a quantitative risk analysis. For each safety indicator, a goal, a bounded interval and proper bases are developed, to allow for a clear and comprehensive individual behavior evaluation. On the establishment of the intervals and boundaries, a probabilistic safety study, operational experience, international and national standards and technical specifications were used. Additionally, an integrated evaluation of the indicators, using expert systems, was done to obtain an overview of the plant general safety. This evaluation uses well-defined and clear rules and weights for each indicator to be considered. These rules were implemented by means of a computational language, on a friendly interface, so that it is possible to obtain a quick response about operational safety. This methodology can be used to identify situations where the plant safety is challenged, by giving a general overview of the plant operational condition. Additionally, this study can

  16. Baseline Evaluations to Support Control Room Modernization at Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Boring, Ronald L.; Joe, Jeffrey C.

    2015-02-01

    For any major control room modernization activity at a commercial nuclear power plant (NPP) in the U.S., a utility should carefully follow the four phases prescribed by the U.S. Nuclear Regulatory Commission in NUREG-0711, Human Factors Engineering Program Review Model. These four phases include Planning and Analysis, Design, Verification and Validation, and Implementation and Operation. While NUREG-0711 is a useful guideline, it is written primarily from the perspective of regulatory review, and it therefore does not provide a nuanced account of many of the steps the utility might undertake as part of control room modernization. The guideline is largely summative—intended to catalog final products—rather than formative—intended to guide the overall modernization process. In this paper, we highlight two crucial formative sub-elements of the Planning and Analysis phase specific to control room modernization that are not covered in NUREG-0711. These two sub-elements are the usability and ergonomics baseline evaluations. A baseline evaluation entails evaluating the system as-built and currently in use. The usability baseline evaluation provides key insights into operator performance using the control system currently in place. The ergonomics baseline evaluation identifies possible deficiencies in the physical configuration of the control system. Both baseline evaluations feed into the design of the replacement system and subsequent summative benchmarking activities that help ensure that control room modernization represents a successful evolution of the control system.

  17. Baseline Evaluations to Support Control Room Modernization at Nuclear Power Plants

    International Nuclear Information System (INIS)

    Boring, Ronald L.; Joe, Jeffrey C.

    2015-01-01

    For any major control room modernization activity at a commercial nuclear power plant (NPP) in the U.S., a utility should carefully follow the four phases prescribed by the U.S. Nuclear Regulatory Commission in NUREG-0711, Human Factors Engineering Program Review Model. These four phases include Planning and Analysis, Design, Verification and Validation, and Implementation and Operation. While NUREG-0711 is a useful guideline, it is written primarily from the perspective of regulatory review, and it therefore does not provide a nuanced account of many of the steps the utility might undertake as part of control room modernization. The guideline is largely summative–intended to catalog final products–rather than formative–intended to guide the overall modernization process. In this paper, we highlight two crucial formative sub-elements of the Planning and Analysis phase specific to control room modernization that are not covered in NUREG-0711. These two sub-elements are the usability and ergonomics baseline evaluations. A baseline evaluation entails evaluating the system as-built and currently in use. The usability baseline evaluation provides key insights into operator performance using the control system currently in place. The ergonomics baseline evaluation identifies possible deficiencies in the physical configuration of the control system. Both baseline evaluations feed into the design of the replacement system and subsequent summative benchmarking activities that help ensure that control room modernization represents a successful evolution of the control system.

  18. Evaluation of Next Generation Nuclear Power Plant (NGNP) Intermediate Heat Exchanger (IHX) Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    E. A. Harvego

    2006-04-01

    This report summarizes results of a preliminary evaluation to determine the operating conditions for the Next Generation Nuclear Plant (NGNP) Intermediate Heat Exchanger (IHX) that will transfer heat from the reactor primary system to the demonstration hydrogen production plant(s). The Department of Energy is currently investigating two primary options for the production of hydrogen using a high temperature reactor as the power source. These options are the High Temperature Electrolysis (HTE) and Sulfur-Iodine (SI) thermochemical hydrogen production processes. However, since the SI process relies entirely on process heat from the reactor, while the HTE process relies primarily on electrical energy with only a small amount of process heat required, the design of the IHX is dictated by the SI process heat requirements. Therefore, the IHX operating conditions were defined assuming 50 MWt is available for the production of hydrogen using the SI process. Three configurations for the intermediate loop were evaluated, including configurations for both direct and indirect power conversion systems. The HYSYS process analysis software was used to perform sensitivity studies to determine the influence of reactor outlet temperatures, intermediate loop working fluids (helium and molten salt), intermediate loop pressures, and intermediate loop piping lengths on NGNP performance and IHX operating conditions. The evaluation of NGNP performance included assessments of overall electric power conversion efficiency and estimated hydrogen production efficiency. Based on these evaluations, recommended IHX operating conditions are defined.

  19. Improving motor reliability in nuclear power plants: Volume 1, Performance evaluation and maintenance practices

    International Nuclear Information System (INIS)

    Subudhi, M.; Gunther, W.E.; Taylor, J.H.; Sugarman, A.C.; Sheets, M.W.

    1987-11-01

    This report constitutes the first of the three volumes under this NUREG. The report presents recommendations for developing a cost-effective program for performance evaluation and maintenance of electric motors in nuclear power plants. These recommendations are based on current industry practices, available techniques for monitoring degradation in motor components, manufacturer's recommendations, operating experience, and results from two laboratory tests on aged motors. Two laboratory test reports on a small and a large motor are presented in separate volumes of this NUREG. These provide the basis for the various functional indicators recommended for maintenance programs in this report. The overall preventive maintenance program is separated into two broad areas of activity aimed at mitigating the potential effects of equipment aging: Performance Evaluation and Equipment Maintenance. The latter involves actually maintaining the condition of the equipment while the former involves those activities undertaken to monitor degradation due to aging. These monitoring methods are further categorized into periodic testing, surveillance testing, continuous monitoring and inspections. This study focuses on the methods and procedures for performing the above activities to maintain the motors operationally ready in a nuclear facility. This includes an assessment of various functional indicators to determine their suitability for trending to monitor motor component condition. The intrusiveness of test methods and the present state-of-the-art for using the test equipment in a plant environment are discussed. In conclusion, implementation of the information provided in this report, will improve motor reliability in nuclear power plants. The study indicates the kinds of tests to conduct, how and when to conduct them, and to which motors the tests should be applied. 44 refs., 12 figs., 13 tabs

  20. Molecular Fingerprinting Approach in Plant Species Evaluation for a Nuclear Power Programme

    International Nuclear Information System (INIS)

    Azhar Mohamed

    2011-01-01

    Deoxyribonucleic acid (DNA) as a tool for marker technology is found to be remarkable, advanced and exciting in recent years. DNA markers are valuable tools and important in various plant breeding analyses for identification, gene mapping, marker systems and mutagenesis response. As gene expression is related to concurrent cellular activities and is mobilised in the adaptation of plants to adverse environmental conditions, changes at the DNA levels can be detected simultaneously. The changes also reflect response onto plant traits in which selection for better quality plant materials can be made and/or used as bio-indicator response in tracking any environmental change. The objective of the present study is to show Inter Simple Sequence Repeat (ISSR) markers as an important technique in differentiating plant DNA genomic in various species for the evaluation of their diversity and radiation effects in population. The technique has been found to be rapid, simple, reliable and robust in generating molecular fingerprinting database in bio surveillance for a nuclear power programme. (author)

  1. Human factors evaluation of man-machine interface for periodic safety review of nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Yong Hee; Lee, Jung Woon; Park, Jae Chang; Hwang, In Koo; Lee, Hyun Cheol; Jang, Tong Il; Ku, Jin Young; Kim, Soo Jin

    2004-12-01

    This report describes the research results of human factors assessment on the MMI(Man Machine Interface) equipment as part of Periodic Safety Review(PSR) of Nuclear Power Plants(NPPs). As MMI is a key factor among human factors to be reviewed in PSR, we reviewed the MMI components of nuclear power plants in aspect of human factors engineering. The availability, suitability, and effectiveness of the MMI devices were chosen to be reviewed. The MMI devices were investigated through the review of design documents related to the MMI, survey of control panels, evaluation of experts, and experimental assessment. Checklists were used to perform this assessment and record the review results. The items mentioned by the expert comments to review in detail in relation with task procedures were tested by experiments with operators' participation. For some questionable issues arisen during this MMI review, operator workload and possibility of errors in operator actions were analysed. The reviewed MMI devices contain MCR(Main Control Room), SPDS(Safety Parameter Display System), RSP(Remote Shutdown Panel), and the selected LCBs(Local Control Boards) importantly related to safety. As results of the assessments, any significant problem challenging the safety was not found on human factors in the MMI devices. However, several small items to be changed and improved in suitability of MMI devices were discovered. An action plan is recommended to accommodate the suggestions and review comments. It will enhance the plant safety on MMI area

  2. Measurement and evaluation of human factor training in nuclear power plants

    International Nuclear Information System (INIS)

    Hamasaki, Kenichi

    2006-01-01

    The purpose of this study is to measure and evaluate the effectiveness of human factor training aimed at awareness and behavioral changes, conducted by electric power company for the nuclear power plant staff. As the first step, the researcher investigated recent trends in training measurement and evaluation methods in the United States. It was found that many instances of training measurement/evaluation had been reported, and that the ROI model was the mainstream method for such measurement and evaluation. However, there had been no instances reported in which the effectiveness of human factor training for plant staff had been measured. The researcher therefore developed a new questionnaire-type of effectiveness measurement/evaluation method, based on the framework of the ROI model. Two-years of research was then conducted, in which the effectiveness of a human factor training program was measured using the newly developed method. This research revealed that participants' overall satisfaction and knowledge/skill acquisition levels were high. The percentage of participants who demonstrated awareness/behavioral change after returning to the workplace increased from 50% at first measurement to 81% at second measurement. It can therefore be concluded that the effectiveness of the second training is greater than that of the first training. Use of the new effectiveness measurement/evaluation method will enable quantification of human factor training effectiveness and help improve training quality. (author)

  3. A safety evaluation of fire and explosion in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Nishio, Gunji; Takada, Junichi; Tukamoto, Michio; Watanabe, Kouji; Miyata, Teijirou

    1996-01-01

    The demonstration test was performed in JAERI to prove the adequacy of a safety evaluation for an air-ventilation system in the case of solvent fire and red-oil explosion in a nuclear fuel reprocessing plant. The test objectives were to obtain data of the safety evaluation on a thermofluid behavior and a confinement effect of radioactive materials during fire and explosion while the system is operating in a cell. The computer code was developed to evaluate the safety of associated network in the ventilation system and to estimate the confinement of radioactive materials in the system. The code was verified by comparison of code calculations with results of the demonstration test. (author)

  4. Applying Human Factors Evaluation and Design Guidance to a Nuclear Power Plant Digital Control System

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Ulrich; Ronald Boring; William Phoenix; Emily Dehority; Tim Whiting; Jonathan Morrell; Rhett Backstrom

    2012-08-01

    The United States (U.S.) nuclear industry, like similar process control industries, has moved toward upgrading its control rooms. The upgraded control rooms typically feature digital control system (DCS) displays embedded in the panels. These displays gather information from the system and represent that information on a single display surface. In this manner, the DCS combines many previously separate analog indicators and controls into a single digital display, whereby the operators can toggle between multiple windows to monitor and control different aspects of the plant. The design of the DCS depends on the function of the system it monitors, but revolves around presenting the information most germane to an operator at any point in time. DCSs require a carefully designed human system interface. This report centers on redesigning existing DCS displays for an example chemical volume control system (CVCS) at a U.S. nuclear power plant. The crucial nature of the CVCS, which controls coolant levels and boration in the primary system, requires a thorough human factors evaluation of its supporting DCS. The initial digital controls being developed for the DCSs tend to directly mimic the former analog controls. There are, however, unique operator interactions with a digital vs. analog interface, and the differences have not always been carefully factored in the translation of an analog interface to a replacement DCS. To ensure safety, efficiency, and usability of the emerging DCSs, a human factors usability evaluation was conducted on a CVCS DCS currently being used and refined at an existing U.S. nuclear power plant. Subject matter experts from process control engineering, software development, and human factors evaluated the DCS displays to document potential usability issues and propose design recommendations. The evaluation yielded 167 potential usability issues with the DCS. These issues should not be considered operator performance problems but rather opportunities

  5. Nuclear power plant emergency preparedness: results from an evaluation of Michigan's potassium iodide distribution program.

    Science.gov (United States)

    Zwolinski, Laura R; Stanbury, Martha; Manente, Susan

    2012-10-01

    In 2009, the Michigan Department of Community Health (MDCH) made potassium iodide (KI), a nonprescription radio-protective drug, available by mailing vouchers redeemable at local pharmacies for KI tablets, at no cost to residents living within 10 miles of Michigan's 3 nuclear power plants (NPPs). MDCH conducted an evaluation of this program to determine Michigan's KI coverage and to assess general emergency preparedness among residents living near the NPPs. KI coverage was estimated based on redeemed voucher counts and the 2010 Census. Telephone surveys were administered to a random sample (N = 153) of residents living near Michigan's NPPs to evaluate general emergency preparedness, reasons for voucher use or nonuse, and KI knowledge. Only 5.3% of eligible residences redeemed KI vouchers. Most surveyed residents (76.5%) were aware of living near an NPP, yet 42.5% reported doing "nothing" to plan for an emergency. Almost half of surveyed voucher users did not know when to take KI or which body part KI protects. Among voucher nonusers, 48.0% were either unaware of the program or did not remember receiving a voucher. Additional efforts are needed to ensure that all residents are aware of the availability of KI and that recipients of the drug understand when and why it should be taken. Minimal emergency planning among residents living near Michigan's NPPs emphasizes the need for increased emergency preparedness and awareness. Findings are particularly salient given the March 2011 Fukushima Daiichi Nuclear Power Plant emergency in Japan.

  6. Design and evaluation of warning systems: application to nuclear power plants

    International Nuclear Information System (INIS)

    Pe Benito-Claudio, C.

    1986-01-01

    This study starts by defining and explaining key concepts about warning, both as a process and a system. Thereafter, it presents a quantitative, probabilistic, and decision-oriented methodology for designing and evaluating a warning system. It illustrates the methodology for the case of rare, controllable, and potentially disastrous technological events, such as accidents in nuclear power plants. The methodology covers and links the three principal components of a warning system - signal (which is mainly technical), warning dissemination, and warning response (which are mainly social) - thereby allowing the relative evaluation of technological and social measures for reducing risks. Analytical principles and techniques of risk and decision analyses are applied. It defines a probabilistic performance measure to characterize each component of a warning system, and a value measure to assess the overall effectiveness of the system. An important aspect of this work is the integration, into one analytical model, of the results of engineering studies, such as probabilistic risk assessments of nuclear power plants, and of empirical findings on human response to warning in sociological research. The models, calculations, and sensitivity analyses are done with influence diagrams that are both intuitive and mathematical. This work puts particular emphasis on the study of behavioral response of individuals to warning

  7. Evaluation of digital fault-tolerant architectures for nuclear power plant control systems

    International Nuclear Information System (INIS)

    Battle, R.E.

    1990-01-01

    Four fault tolerant architectures were evaluated for their potential reliability in service as control systems of nuclear power plants. The reliability analyses showed that human- and software-related common cause failures and single points of failure in the output modules are dominant contributors to system unreliability. The four architectures are triple-modular-redundant (TMR), both synchronous and asynchronous, and also dual synchronous and asynchronous. The evaluation includes a review of design features, an analysis of the importance of coverage, and reliability analyses of fault tolerant systems. An advantage of fault-tolerant controllers over those not fault tolerant, is that fault-tolerant controllers continue to function after the occurrence of most single hardware faults. However, most fault-tolerant controllers have single hardware components that will cause system failure, almost all controllers have single points of failure in software, and all are subject to common cause failures. Reliability analyses based on data from several industries that have fault-tolerant controllers were used to estimate the mean-time-between-failures of fault-tolerant controllers and to predict those failures modes that may be important in nuclear power plants. 7 refs., 4 tabs

  8. Nuclear Power Plants. Revised.

    Science.gov (United States)

    Lyerly, Ray L.; Mitchell, Walter, III

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

  9. Development of an integrated database management system to evaluate integrity of flawed components of nuclear power plant

    International Nuclear Information System (INIS)

    Mun, H. L.; Choi, S. N.; Jang, K. S.; Hong, S. Y.; Choi, J. B.; Kim, Y. J.

    2001-01-01

    The object of this paper is to develop an NPP-IDBMS(Integrated DataBase Management System for Nuclear Power Plants) for evaluating the integrity of components of nuclear power plant using relational data model. This paper describes the relational data model, structure and development strategy for the proposed NPP-IDBMS. The NPP-IDBMS consists of database, database management system and interface part. The database part consists of plant, shape, operating condition, material properties and stress database, which are required for the integrity evaluation of each component in nuclear power plants. For the development of stress database, an extensive finite element analysis was performed for various components considering operational transients. The developed NPP-IDBMS will provide efficient and accurate way to evaluate the integrity of flawed components

  10. Evaluation of the safety of the operating nuclear power plants built to earlier standards

    International Nuclear Information System (INIS)

    Menteseoglu, S.

    2001-01-01

    The objective of this paper is to provide practical assistance on judging the safety of a nuclear power plant, on the basis of a comparison with current safety standards and operational practices. For nuclear power plants built to earlier standards for which there are questions about the adequacy of the maintenance of the plant design and operational practices, a safety review against current standards and practices can be considered a high priority. The objective of reviewing nuclear power plants built to earlier standards against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The safety significance of the issues identified should be judged according to their implications for plant design and operation in terms of basic safety concepts such as defence in depth and safety culture. In addition, this paper provides assistance on the prioritization of corrective measures and their implementation so as to approach an acceptable level of safety

  11. EARLY-STAGE DESIGN AND EVALUATION FOR NUCLEAR POWER PLANT CONTROL ROOM UPGRADES

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring; Jeffrey C. Joe; Thomas A. Ulrich; Roger T. Lew

    2015-03-01

    As control rooms are modernized with new digital systems at nuclear power plants, it is necessary to evaluate operator performance with these systems as part of a verification and validation process. While there is regulatory and industry guidance for some modernization activities, there are no well defined standard processes or predefined metrics available for assessing what is satisfactory operator interaction with new systems, especially during the early design stages. This paper proposes a framework defining the design process and metrics for evaluating human system interfaces as part of control room modernization. The process and metrics are generalizable to other applications and serve as a guiding template for utilities undertaking their own control room modernization activities.

  12. Licensing evaluation of CANDU-PHW nuclear power plants relative to U.S. regulatory requirements

    International Nuclear Information System (INIS)

    Erp, J.B. van

    1978-01-01

    Differences between the U.S. and Canadian approach to safety and licensing are discussed. U.S. regulatory requirements are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to current Regulatory Requirements and Guides. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S. These modifications are proposed solely for the purpose of maintaining consistency within the current U.S. regulatory system and not out of a need to improve the safety of current-design CANDU-PHW nuclear power plants. A number of issues are identified which still require resolution. Most of these issues are concerned with design areas not (yet) covered by the ASME code. (author)

  13. Means of evaluating and improving the effectiveness of training of nuclear power plant personnel

    International Nuclear Information System (INIS)

    2003-07-01

    In 1996 the IAEA published a guidebook on Nuclear Power Plant Personnel Training and its Evaluation, which constitutes the recommendations of the IAEA with respect to development, implementation and evaluation of training programmes. The IAEA Technical Working Group on Training and Qualification of Nuclear Power Plant Personnel recommended that an additional publication be prepared to provide further details concerning the effectiveness of NPP personnel training. As the nuclear power industry continues to be challenged by increasing safety requirements, a high level of competition and decreasing budgets, it becomes more important than ever to have some methodology of ensuring that training contributes a value to the organization. The actual determination of training effectiveness is not an easy task because of the many variables associated with personnel performance. For example, for training to make a difference in job performance, line management should be involved prior to training delivery to identify what performance is desired, but not being achieved. Then, training is developed to meet desired performance, which is followed by practice and continued management reinforcement. Because of these other variables, it is very difficult to prove that training had a sole contribution to performance improvement, but rather one of many contributors needed for performance improvement. The difficulty to isolate training as a sole contributor has been documented in a number of research studies over the recent years. Due to these limitations, a base assumption must be made in order to use any methodology for training effectiveness evaluation. That assumption is that there are some basic principles for developing training and if training programmes are developed and maintained using these principles, then the training provided should be an effective tool to improve the line organization performance. By monitoring various types of training effectiveness indications

  14. Evaluation of external hazards to nuclear power plants in the United States

    International Nuclear Information System (INIS)

    Kimura, C.Y.; Budnitz, R.J.

    1987-12-01

    The Lawrence Livermore National Laboratory (LLNL) has performed a study of the risk of core damage to nuclear power plants in the United States due to externally initiated events. The broad objective has been to gain an understanding of whether or not each external initiator is among the major potential accident initiators that may pose a threat of severe reactor core damage or of large radioactive release to the environment from the reactor. Four external hazards were investigated in this report. These external hazards are internal fires, high winds/tornadoes, external floods, and transportation accidents. Analysis was based on two figures-of-merit, one based on core damage frequency and the other based on the frequency of large radioactive releases. Using these two figures-of-merit as evaluation criteria, it has been feasible to ascertain whether the risk from externally initiated accidents is, or is not, an important contributor to overall risk for the US nuclear power plants studied. This has been accomplished for each initiator separately. 208 refs., 17 figs., 45 tabs

  15. Procedures for evaluation of vibratory ground motions of soil deposits at nuclear power plant sites

    International Nuclear Information System (INIS)

    1975-06-01

    According to USNRC requirements set forth in Appendix A, 10 CFR, Part 100, vibratory ground motion criteria for a nuclear plant must be based on local soil conditions, as well as on the seismicity, geology, and tectonics of the region. This report describes how such criteria can be developed by applying the latest technology associated with analytical predictions of site-dependent ground motions and with the use of composite spectra obtained from the current library of strong motion records. Recommended procedures for defining vibratory ground motion criteria contain the following steps: (1) geologic and seismologic studies; (2) site soils investigations; (3) site response sensitivity studies; (4) evaluation of local site response characteristics; (5) selection of site-matched records; and (6) appraisal and selection of seismic input criteria. An in-depth discussion of the engineering characteristics of earthquake ground motions including parameters used to characterize earthquakes and strong motion records, geologic factors that influence ground shaking, the current strong motion data base, and case histories of the effects of past earthquake events is presented. Next, geotechnical investigations of the seismologic, geologic, and site soil conditions required to develop vibratory motion criteria are briefly summarized. The current technology for establishing vibratory ground motion criteria at nuclear plant sites, including site-independent and site-dependent procedures that use data from strong motion records and from soil response analyses is described. (auth)

  16. Study on the analytical methods to evaluate the interaction effects of embedded nuclear power plants

    International Nuclear Information System (INIS)

    Fukuzawa, R.; Chiba, O.; Tohdo, M.; Hatori, T.

    1987-01-01

    For the seismic response analyses of nuclear power plants which are more or less embedded into soil medium, it is very important to evaluate the interactive effects on the interface between soil and foundation wall, such as wave propagation to infinity radiated from the side wall. The methods generally used in these analyses can be classified into the two basic procedures. One is the direct procedure in which a structure and soil are medeled together into finite elements or beam elements having the quiet boundary; transmitting boundary, non-reflecting boundary. The other is the substructure procedure in which the computational procedure is divided into two steps. In the first step the behavior of massless foundation due to an external force and seismic motion under the assumption as rigid foundation is analyzed, and the results are represented by impedance functions and foundation input motions. Using the results, the structural response analyses are performed. Taking the substructure procedure for seismic response analyses of embedded nuclear power plants, it is the purpose of this paper to present the features of solutions on impedance functions and foundation input motions obtained by various methods such as 3- or 2-dimensional approach, and their influences on structural responses

  17. Methods and benefits of experimental seismic evaluation of nuclear power plants. Final report

    International Nuclear Information System (INIS)

    1979-07-01

    This study reviews experimental techniques, instrumentation requirements, safety considerations, and benefits of performing vibration tests on nuclear power plant containments and internal components. The emphasis is on testing to improve seismic structural models. Techniques for identification of resonant frequencies, damping, and mode shapes, are discussed. The benefits of testing with regard to increased damping and more accurate computer models are oulined. A test plan, schedule and budget are presented for a typical PWR nuclear power plant

  18. Methods and benefits of experimental seismic evaluation of nuclear power plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-01

    This study reviews experimental techniques, instrumentation requirements, safety considerations, and benefits of performing vibration tests on nuclear power plant containments and internal components. The emphasis is on testing to improve seismic structural models. Techniques for identification of resonant frequencies, damping, and mode shapes, are discussed. The benefits of testing with regard to increased damping and more accurate computer models are oulined. A test plan, schedule and budget are presented for a typical PWR nuclear power plant.

  19. Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2

    International Nuclear Information System (INIS)

    1995-09-01

    This report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985). Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), Supplement No. 8 (January 1992). Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April 1993), Supplement No. 12 (October 1993). Supplement No. 13 (April 1994), Supplement No. 14 (December 1994), and Supplement No. 15 (June 1995) issued by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos, 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER

  20. A Human Reliability Evaluation Tool for Main Control Rooms in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Kim, Jae Whan; Won Dae

    2012-01-01

    Identification and quantification of main control room (MCR) operators' errors are important for preventing undesired situations and enhancing reliability of nuclear power plants (NPPs). Human errors could be predicted by human error evaluation methods and appropriate interface design and training/education programs can be made based on the evaluation results. The accident sequence evaluation program (ASEP) and technique for human error rate prediction (THERP) methods have been mainly used for the human reliability analysis (HRA) in the probabilistic safety assessment (PSA) of Korean NPPs. The operational environments of these computerized MCRs (advanced MCRs) are totally different from those of conventional MCRs. The different interfaces require operators to perform different tasks for operating and maintaining plants. While only primary tasks are considered in conventional MCRs, secondary tasks such as interface managements are one of the major concerns of advanced MCRs. This computerized operational environment may make operation tasks more convenient but may make some controversial issues of human errors or cause new types of human errors. For the changed interfaces, different design related influencing factors (DIFs) should be considered according to the design characteristics. KAERI (Korea Atomic Energy Research Institute) developed a standard HRA method for the PSA of Korean NPPs. It is based on the ASEP and THERP methods. For the advanced MCRs, the modified HRA method was proposed by KAERI based on the basic model of the HRA for conventional MCRs. The newly identified DIFs have been reflected into the conventional HRA framework as performance shaping factors (PSFs)

  1. Reload safety evaluation report for kori nuclear power plant unit 2 cycle 9

    International Nuclear Information System (INIS)

    Cho, Beom Jin; Kim, Si Yong; Kim, Oh Hwan; Nam, Kee Il; Um, Gil Sup; Ban, Chang Hwan; Choi, Dong Uk; Yoon, Kyung Ho

    1992-04-01

    The Kori Nuclear Power Plant Unit 2 (Kori-2) is anticipated to be refuelled with 16x16 Korean Fuel Assemblies (KOFA), which are based on the KAERI design starting from Cycle 8. This report presents a reload safety evaluation for Kori-2, Cycle 9 and demonstrates that the reactor core being composed of various fuel assembly types as described below will not adversely affect the safety of the public and the plant. The evaluation of Kori-2, Cycle 9 was accomplished utilizing the methodology described in 'Reload Transition Safety Report for KORI 2' (Ref. /1-1/). The reload core for Kori-2, Cycle 9 is entirely comprised of 16x16 KOFA. In the Kori-2 licensing documentation to KEPCO the reference safety evaluation was provided for the operation of a reactor core fully loaded with KOFA as well as associated proposed changes to the Kori-2 Technical Specifications. The reload for Kori-2, Cycle 9 also introduces UO 2 /Gd 2 O 3 containing fuel rods. The use of fuel rods with Gd 2 O 3 poisoning of the fuel has been approved as a part of the above mentioned licensing documentation. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 9 core design described herein. (Author)

  2. Nuclear power plant siting

    International Nuclear Information System (INIS)

    Sulkiewicz, M.; Navratil, J.

    The construction of a nuclear power plant is conditioned on territorial requirements and is accompanied by the disturbance of the environment, land occupation, population migration, the emission of radioactive wastes, thermal pollution, etc. On the other hand, a nuclear power plant makes possible the introduction of district heating and increases the economic and civilization activity of the population. Due to the construction of a nuclear power plant the set limits of negative impacts must not be exceeded. The locality should be selected such as to reduce the unfavourable effects of the plant and to fully use its benefits. The decision on the siting of the nuclear power plant is preceded by the processing of a number of surveys and a wide range of documentation to which the given criteria are strictly applied. (B.H.)

  3. Visual Fatigue Evaluation: Improvement of Reflected Glare on Touch Screen for Nuclear Power Plant

    International Nuclear Information System (INIS)

    Liu, Kang Hung; Yang, Chih Wei; Hwang, Sheue Ling; Liou, Jin Liang

    2014-01-01

    The purpose of this research is to evaluate the visual fatigue of operators caused by glare problems in the main control room of nuclear power plant. Within the limitation in the main control room, reflectors were set under the light source which generates reflected glare on touch screens. Through avoiding the light directly shines on touch screens, reflected glare were eliminated. This research matched up the setting process of reflectors, evaluated the visual fatigue of operators, and collected user's opinions before reflector setting, after the first setting, and after the second setting. The design of reflectors could refer the result of evaluations and the collection of opinions. Nevertheless, the improvement of reflected glare on touch screens could be verified by this evaluations. The result showed that setting reflectors under the light source could eliminate reflected glare effectively, and the visual fatigue was reduced both on subject and object evaluations. However, the setting direction of reflectors has potential effect on operators' visual fatigue, so the real setting of reflectors still need to be evaluated completely. The near point accommodation could reflect the effect of visual fatigue caused by changes of lighting environment. Thus, the verification of new lighting environment according to the near point accommodation is suggested

  4. Atucha I nuclear power plant transients analysis

    International Nuclear Information System (INIS)

    Castano, J.; Schivo, M.

    1987-01-01

    A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

  5. Seismic evaluation of non-seismically designed existing Magnox nuclear power plants

    International Nuclear Information System (INIS)

    Kunar, R.R.

    1984-01-01

    The philosophy and method adopted for the seismic assessment of three existing Magnox nuclear stations in the United Kingdom are presented in this paper. The plants were not seismically designed. The particular procedures that were applied were tailored to suit the difficulties of lack of data which is somewhat inevitable for plants designed and built about 25 to 30 years ago. Special procedures included on-site testing with a portable shake table, low vibration testing using a structural dynamics analyser, and on-site inspections. The low vibration testing was most invaluable in detecting differences between 'as-built' conditions and the engineering drawings. From the point of view of economics, this was more effective than conducting full structural surveys to determine the as-built conditions. The testing results also provided confidence in the answers from numerical models. The philosophy adopted for the Magnox reactors in the seismic assessment was to determine what peak ground accelerations the sites can sustain and then evaluate the chances of exceeding the ground accelerations over the remaining lifetime of the plants. The peak ground acceleration for each site was determined on the basis of the criteria of safe shutdown and prevention of significant off-site radiological exposure

  6. Feasibility of quantitative performance measures for evaluating nuclear power plant operators

    International Nuclear Information System (INIS)

    Carter, R.J.; Connelly, E.M.; Krois, P.A.

    1989-01-01

    A more valid measure of team performance in nuclear power plants is needed. A study is described which was oriented towards evaluating the feasibility of synthesizing performance measures by deriving measures for crews responding to an off-normal event in a full-scope simulator. The thesis was that performance assessment is based on the subjective judgment of training instructors. The procedure used to synthesize the performance measure consisted of: identification of the factors believed to be important to performance assessment, development of example crew performances and ratings on each by instructors, and derivation of the measure by capturing the instructors' assessment rules. A performance measure was derived which explains nearly all of the variance of the instructors' team performance assessments. There is reason to believe that this method of synthesizing measures can be applied to other events. 7 tabs

  7. Evaluation of environmental tritium level in preoperational period of Cernavoda CANDU Nuclear Power Plant

    International Nuclear Information System (INIS)

    Paunescu, N.; Cotarlea, M.; Galeriu, D.; Margineanu, R.; Mocanu, N.

    1999-01-01

    In Romania, a CANDU Nuclear Power Plant with five reactors of 600 MWe is under construction at Cernavoda town and in 1996 the first reactor was put in operation. The background level of tritium concentration was determined in preoperational stage. The mean values determined are: (7.4±5.5) Bq/l in air humidity, (3.1±1.0) Bq/l in water, (3.5±0.7) Bq/l in tissue water from vegetable and (4.9±1.7) Bq/l in tissue water from cereals. The transfer parameters for deposition from atmosphere on forage and crops (P 14 ), and the contamination of land and vegetation by spray irrigation water, (P 24 ) were evaluated at (29-49) and (0.90±0.27), respectively. (author)

  8. Team interaction skills evaluation criteria for nuclear power plant control room operators

    International Nuclear Information System (INIS)

    Montgomery, J.C.; Toquam, J.; Gaddy, C.

    1991-09-01

    Previous research has shown the value of good team interaction skills to group performance, yet little progress has been made on in terms of how such skills can be measured. In this study rating scales developed previously (Montgomery, et al., 1990) were extensively revised and cast into a Behaviorally Anchored Rating Scale (BARS) and a Behavioral Frequency format. Rating data were collected using 13 training instructors at the Diablo Canyon Nuclear Plant, who rated three videotapes of simulator scenario performance during a day-long training session and later evaluated control room crews during requalification training. High levels of interrater agreement on both rating scales were found. However, the factor structure of the ratings was generally inconsistent with that hypothesized. Analysis of training ratings using Cronbach's components of accuracy (Cronbach, 1955) indicated that BARS ratings generally exhibited less error than did the Behavioral Frequency ratings. The results are discussed in terms of both field and research implications

  9. Evaluation of ion exchange resin performance in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Wen, T.J.; Chen Huang, T.; Liu, W.C.; Liu, T.C.; Shen, S.C.

    2004-01-01

    Different types and brands of bead and powdered ion exchange resin have been widely used in condensate and reactor water clean-up system in Boiling Water Reactors (BWRs). The importance of primary coolant quality control has been well recognized. Recently, some measures both in promoting the conductivity of reactor water and minimizing iron crud of feed-water have been recommended by the Institute of Nuclear Energy Research. The variation of water quality between the condensate and reactor water is closely related with the performance of condensate polisher and precoat condition. In order to ensure optimization of water chemistry quality in reactor system, historical water chemistry data have been reviewed and evaluated. The most probable causes of exceeding the chemistry limits and guidelines should be identified and some reasonable corrective actions have been documented as a reference for promoting the water quality and run length extending in the plant normal operation. (author)

  10. Seismic analysis and structure capacity evaluation of the Belene nuclear power plant

    International Nuclear Information System (INIS)

    Johnson, J.J.; Hashimoto, P.S.; Campbell, R.D.; Baltus, R.S.

    1993-01-01

    The seismic analysis and structure capacity evaluation of the Belene Nuclear Power Plant, a two-unit WWER 1000, was performed. The principal objective of the study was to review the major aspects of the seismic design including ground motion specification, foundation concept and materials, and the Unit I main reactor building structure response and capacity. The main reactor building structure /foundation/soil were modeled and analyzed by a substructure approach to soil-structure interaction (SSI) analysis. The elements of the substructure approach, implemented in the family of computer programs CLASSI, are: Specification of the free-field ground motion; Modeling the soil profile; SSI parameters; Modeling the structure; SSI-response analyses. Each of these aspects is discussed. The Belene Unit 1 main reactor building structure was evaluated to verify the seismic design with respect to current western criteria. The structural capacity evaluation included criteria development, element load distribution analysis, structural element selection, and structural element capacity evaluation. Equipment and commodity design criteria were similarly reviewed and evaluated. Methodology results and recommendations are presented. (author)

  11. Criteria for seismic evaluation and potential design fixes for WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-01-01

    The purpose for this document is to provide a criteria for the seismic evaluation and development of potential design fixes for structures, systems and components for the WWER type Nuclear power plants. The design fixes are divided into two categories, detailed and easy fixes. Detailed fixes are typically applicable to building structures, components for which there is little or no seismic capacity information, large tanks and vital systems and components which make up the reactor cooling system and components which perform support or auxiliary functions. In case of the design of 'easy fixes', the criteria presented may be used for both the seismic design as well as for the evaluation of structures, systems and components to which easy fix design applies. Easy fixes are situations where seismic capacities of structures, systems and components can be significantly increased with relatively minor, inexpensive fixes usually associated with anchorage modification of safety related structures, systems and components or those that could interact with safety related structures, systems and components. Often these fixes can be accomplished while the plant is in operation

  12. External human induced events in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The purpose of the present Safety Guide is to provide recommendations and guidance for the examination of the region considered for site evaluation for a plant in order to identity hazardous phenomena associated with human induced events initiated by sources external to the plant. In some cases it also presents preliminary guidance for deriving values of relevant parameters for the design basis. This Safety Guide is also applicable for periodic site evaluation and site evaluation following a major human induced event, and for the design and operation of the site's environmental monitoring system. Site evaluation includes site characterization. Consideration of external events that could lead to a degradation of the safety features of the plant and cause a release of radioactive material from the plant and/or affect the dispersion of such material in the environment. And consideration of population issues and access issues significant to safety (such as the feasibility of evacuation, the population distribution and the location of resources). The process of site evaluation continues throughout the lifetime of the facility, from siting to design, construction, operation and decommissioning. The external human induced events considered in this Safety Guide are all of accidental origin. Considerations relating to the physical protection of the plant against wilful actions by third parties are outside its scope. However, the methods described herein may also have some application for the purposes of such physical protection. The present Safety Guide may also be used for events that may originate within the boundaries of the site, but from sources which are not directly involved in the operational states of the nuclear power plant units, such as fuel depots or areas for the storage of hazardous materials for the construction of other facilities at the same site. Special consideration should be given to the hazardous material handled during the construction, operation and

  13. External human induced events in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of the present Safety Guide is to provide recommendations and guidance for the examination of the region considered for site evaluation for a plant in order to identity hazardous phenomena associated with human induced events initiated by sources external to the plant. In some cases it also presents preliminary guidance for deriving values of relevant parameters for the design basis. This Safety Guide is also applicable for periodic site evaluation and site evaluation following a major human induced event, and for the design and operation of the site's environmental monitoring system. Site evaluation includes site characterization. Consideration of external events that could lead to a degradation of the safety features of the plant and cause a release of radioactive material from the plant and/or affect the dispersion of such material in the environment. And consideration of population issues and access issues significant to safety (such as the feasibility of evacuation, the population distribution and the location of resources). The process of site evaluation continues throughout the lifetime of the facility, from siting to design, construction, operation and decommissioning. The external human induced events considered in this Safety Guide are all of accidental origin. Considerations relating to the physical protection of the plant against wilful actions by third parties are outside its scope. However, the methods described herein may also have some application for the purposes of such physical protection. The present Safety Guide may also be used for events that may originate within the boundaries of the site, but from sources which are not directly involved in the operational states of the nuclear power plant units, such as fuel depots or areas for the storage of hazardous materials for the construction of other facilities at the same site. Special consideration should be given to the hazardous material handled during the construction, operation and

  14. Nuclear power plant construction

    International Nuclear Information System (INIS)

    Lima Moreira, Y.M. de.

    1979-01-01

    The legal aspects of nuclear power plant construction in Brazil, derived from governamental political guidelines, are presented. Their evolution, as a consequence of tecnology development is related. (A.L.S.L.) [pt

  15. Hypothesis of a nuclear accident to the nuclear power plant of Gravelines with important radioactive release out of the site: risks prevention, intervention strategies. Evaluation of the sensitization to the nuclear risk of the physician practicing near the site

    International Nuclear Information System (INIS)

    Mraovic, Th.

    1998-01-01

    This thesis has for hypothesis a nuclear accident at the nuclear power plant of Gravelines with radioactive release out of the site: the risks prevention and the strategies of intervention are studied. An evaluation of the sensitization to a nuclear risk is made for the general practitioner that practices near the site. (N.C.)

  16. Evaluation of Monticello Nuclear Power Plant, Environmental Impact Prediction, based on monitoring programs

    Energy Technology Data Exchange (ETDEWEB)

    Gore, K.L.; Thomas, J.M.; Kannberg, L.D.; Watson, D.G.

    1976-11-01

    This report evaluates quantitatively the nonradiological environmental monitoring programs at Monticello Nuclear Generating Plant. The general objective of the study is to assess the effectiveness of monitoring programs in the measurement of environmental impacts. Specific objectives include the following: (1) Assess the validity of environmental impact predictions made in the Environmental Statement by analysis of nonradiological monitoring data; (2) evaluate the general adequacy of environmental monitoring programs for detecting impacts and their responsiveness to Technical Specifications objectives; (3) assess the adequacy of preoperational monitoring programs in providing a sufficient data base for evaluating operational impacts; (4) identify possible impacts that were not predicted in the environmental statement and identify monitoring activities that need to be added, modified or deleted; and (5) assist in identifying environmental impacts, monitoring methods, and measurement problems that need additional research before quantitative predictions can be attempted. Preoperational as well as operational monitoring data were examined to test the usefulness of baseline information in evaluating impacts. This included an examination of the analytical methods used to measure ecological and physical parameters, and an assessment of sampling periodicity and sensitivity where appropriate data were available.

  17. Evaluation technology of radiation resistant materials in nuclear power plant and satellite

    International Nuclear Information System (INIS)

    Kang, P. H.; Kim, K. Y.; Lee, C.; Jeong, S. H.; Kim, J. A.

    2007-06-01

    The developed evaluation method using dielectric relaxation characteristic is measuring the increment of dipoles and mobile ions inside PEEK with deterioration, and calculate the dielectric relaxation intensity. Dielectric relaxation intensity were evaluated using Cole-Cole's circular arcs from the multi-frequencies measurements around glass transition temperature. The other proposed method using mechanical relaxation characteristic is measuring the brittleness of the deteriorated PEEK, non-destructively. The values of dielectric relaxation intensity showed the deterioration levels of each aged PEEK. Similarly, mechanical relaxation intensity showed decreasing tendency as increasing ageing time, but it did not show a certain tendency for the multi-aged PEEK. The novel developed evaluation method using dielectric relaxation characteristic will be applicable to classification of EQ, and contribute to the reliability of the lifetime extension in the NPP. These results establish a test method for both TID (Total Ionizing Dose) and SEEs through the evaluation, test, design technology. The results turned out to be quite successful, and these can be applied for parts localization for the nuclear power plant and the satellite

  18. Mathematical model use for evaluation of radioactivity spreading in nuclear power plant

    International Nuclear Information System (INIS)

    Kubik, I.; Gladki, Eh.; Yanchik, O.

    1976-01-01

    On the basis of knowledges of radioactive products behaviour and their spreading in nuclear power plant under normal and accident conditions a KOMPLEX program is developed in the FORTRAN 4 language, permitting to calculate the activity in separate parts of the nuclear power plant with WWR type reactor. The COMPLEX program includes the following subprograms: AZ - PRIM - for estimating active products in fuel, coolant, on the surfaces of fuel element cans and the primary circuit. The subprogram permits to estimate the coolant activity at the expense of fission fragments for 4 different leakage mechanisms: due to diffusion, considerable fuel element damage, contamination of fuel element can surface and fuel washout by coolant; KOR - the program for estimating active corrosion products; ACT - the program for estimating the activity of activation products; CONT - the program for estimating the activity in the nuclear power plant premises (protection envelop) and ventilating pipe. The desciption of the above subprograms is given. For testing of the mathematical model applicability and the possibilities of the corresponding programs the checking calculations for operating parameters of nuclear power plant with WWR type reactor were carried out. The calculation results obtained have shown the applicability of the model suggested and the corresponding programes for nuclear power plant under normal operation and accident conditions [ru

  19. Evaluation of environmental tritium level in the area of Cernavoda Nuclear Power Plant

    International Nuclear Information System (INIS)

    Paunescu, N.; Cotarlea, M.; Galeriu, D.; Margineanu, R.; Mocanu, N.

    1997-01-01

    This work is a continuation of the study entitled E valuation of environmental tritium levels before and after start of a new tritium source for model verification' that commenced two years ago as part of an European research contract having as objective the 'Investigations and Modelling of the Dynamics of Environmental HT/HTO/OBT Levels Resulting from the Tritium Releases'. The aim of this study was to evaluate the environmental tritium level in pre-operational stage of Cernavoda Nuclear Power Plant and the dynamics of variation of tritium concentration in the first operational years of NPP, for determination of site-specific transfer parameters and model validation. Representative samples for Cernavoda area were analyzed; - air humidity; - water from Danube River, Danube-Black Sea Canal and lakes; - drinking, rain and snow water; - HTO concentration in soils at different depths; - tissue free water tritium in vegetal and animal foodstuffs relevant for human diets like cereals (wheat, maize, barley). vegetables (potatoes, tomato, cabbage, onion, bean), grapes and wine. The mean concentration of tritium obtained in 1996 for these types of measurements is presented. The values of tritium concentration in air, water, soil and plants are as low as for areas without contaminating sources. The operation in 1996 of the Cernavoda NPP did not modify the tritium environmental level. (authors)

  20. Work responsability evaluation of a nuclear power plant operator during an emergency event

    International Nuclear Information System (INIS)

    Silva, Nilo Garcia da

    2002-01-01

    In this work, a method to evaluate the Man - Machine Interface, using the Model of Human Process utilized by the Combustion Engineer-CE was used in the Licensing Process of a Nuclear Power Plant, applying methodology in the emergency procedure of a SGTR for Angra 1 NPP. The time needed for the operators to perform the tasks step by step in this procedure was calculated. This time was compared to that found in the GOMS model, by which it is possible to perform an evaluation of workload for this accident. The criteria of the Standard ANSI/58.8-1994 was used to establish the reaction and execution times for these tasks. Practical measures in a Angra 2 NPP full scope simulator were done some steps of the SGTR emergency procedure. The time evaluation in the GOMS model has identified that three was a concentration of tasks, with the execution time near the limits regulated in the Standard above as compared with the times from the CE methodology. Finally, after a comparison of the results from Angra 1 and Angra 2, recommendations were done to the Angra 1 NPP utility in order to improve these identified points. (author)

  1. Seismic PSA. Evaluation of nuclear power plant building and ground and recent earthquake examples

    International Nuclear Information System (INIS)

    Nakamura, Susumu

    2008-01-01

    Relation between seismic design and seismic PSA, and seismic fragility are described. Kashiwazaki-Kariwa Nuclear Power Plant was damaged by the Niigata-ken Chuetsu-oki Earthquake in 2007. The seismic design of this plant and the values of the maximum acceleration observed at the control point were compared. The maximum accelerations of Kashiwazaki-Kariwa 1 to 6 showed larger values than the design values. However, main important buildings were not damaged and three necessary conditions, cut off, cooling and containment, are satisfied. The standard seismic ground motion Ss of seismic design, and restoration and reoperation of nuclear power plant are important. Flow of seismic design, estimation of input seismic ground motion using the standard seismic ground motion, the analytical model of nuclear power plant building, image of real proof stress and response, fire of transformer in Kashiwazaki-Kariwa 3, leakage of water with radioactive substances, subsidence of ground, comparison between acceleration response spectrum and standard seismic ground motion at Onagawa and Shiga Nuclear Power Plant, the fault model, the deep ground structure model, the wave velocities observed by some earthquakes are illustrated. (S.Y.)

  2. Evaluation of the degradation of the service water system in nuclear plants

    International Nuclear Information System (INIS)

    Salaices A, E.

    2003-01-01

    The service water system, the circulation water system, the cooling water system and the protection against fires system so much in nuclear plants as in fossils plants they are being degraded by a wide variety of mechanisms. These mechanisms include microbiologically influenced corrosion, cavitation, erosion-corrosion, erosion by solid particles, corrosion in cracks, stings, general corrosion, galvanic corrosion, sedimentation and obstructions and incrustations in the heat exchangers. In the last years were developed predictive models for the more common degradation forms and were installed in a new application of the CHECWORKS TM code called Cooling Water Application (CWA). This application of the code provides a new technology that so much nuclear facilities as fossil ones can use to modelling specific systems and to carry out corrosion predictions in each one of its components. Presently work the results of the employment of the CHECWORKS CWA code are described to carry out predictions of 12 different corrosion mechanisms that affect to the service water system of a nuclear plant, as well as the recommendations and options that the plant can to consider to reduce indexes of damages. This work can be used for to optimize inspections to the service water system and it gives the bases for similar changes in other nuclear plants. (Author)

  3. Nuclear Power Plants (Rev.)

    Energy Technology Data Exchange (ETDEWEB)

    Lyerly, Ray L.; Mitchell III, Walter [Southern Nuclear Engineering, Inc.

    1973-01-01

    Projected energy requirements for the future suggest that we must employ atomic energy to generate electric power or face depletion of our fossil-fuel resources—coal, oil, and gas. In short, both conservation and economic considerations will require us to use nuclear energy to generate the electricity that supports our civilization. Until we reach the time when nuclear power plants are as common as fossil-fueled or hydroelectric plants, many people will wonder how the nuclear plants work, how much they cost, where they are located, and what kinds of reactors they use. The purpose of this booklet is to answer these questions. In doing so, it will consider only central station plants, which are those that provide electric power for established utility systems.

  4. System approach for evaluation of air pollution toxic compounds in the 30-km area of nuclear power plants

    International Nuclear Information System (INIS)

    Shevtsova, O.V.; Zhigunova, L.N.; Makovskaya, N.A.; Pavlovich, E.L.

    2012-01-01

    The article shows the importance of a systematic approach to address environmental problems that arise during the construction of nuclear power plants, and identified the need to consider the transformation and biotransformation of primary pollutants and monitoring secondary pollutants. The basic pathways of pollutants in the air a 30-km zone of nuclear power plants established. The content of primary and secondary air pollutants identified. The evaluation of general toxic risk from primary pollutant and the calculation of the carcinogenic risk of secondary pollutants entering the body by inhalation are carried out. (authors)

  5. Dose evaluation model for radionuclides released from the spent nuclear fuel reprocessing plant in Rokkasho

    International Nuclear Information System (INIS)

    Hisamatsu, Shun'ichi; Iyogi, Takashi; Inaba, Jiro; Chiang, Jing-Hsien; Suwa, Hiroji; Koide, Mitsuo

    2007-01-01

    A dose evaluation model was developed for radionuclides released from the spent nuclear fuel reprocessing plant which is located in Rokkasho, Aomori Prefecture, and now undergoing test operation. The dose evaluation model suitable for medium- and long-term dose assessments for both prolonged and short-term releases of radionuclides to the atmosphere was developed on the PC. The ARAC-2, a particle tracing type dispersion model coupled with 3-D wind field calculation by a mass conservative model, was adopted as the atmospheric dispersion model. The terrestrial transfer model included movement in soil and groundwater as well as an agricultural and livestock farming system. The available site-specific social and environmental characteristics were incorporated in the model. Growing of the crops was also introduced and radionuclides absorbed were calculated from weight increase from the start of deposition to harvest, and transfer factors. Most of the computer code system of the models was completed by 2005, and this paper reports the results of the development. (author)

  6. Assessment of drinking water quality around Kudankulam nuclear power plant site using fuzzy synthetic evaluation

    International Nuclear Information System (INIS)

    Ramesh, S.; Pratheeba, V.; Murugesan, A.G.; Dahiya, S.

    2007-01-01

    A method based on concept of fuzzy set theory is used for decision-making in the assessment of physicochemical quality of drinking water. Conventional method for water quality assessment does not consider the uncertainties involved either in measurement of water quality parameters or in the limits provided by the regulatory bodies. Fuzzy synthetic evaluation model gives the certainty levels for the quality class of the water based on the prescribed limit of various regulatory bodies and opinion of the experts from the field of drinking water quality. In this paper application of fuzzy rule based method is illustrated with twelve drinking water samples from the residential locality in the vicinity of Kudankulam Nuclear Power Plant site. These samples were analysed for fifteen different physico-chemical parameters, out of them eleven important parameters were used for the quality assessment using fuzzy synthetic evaluation approach. From this study. it has been concluded that out of 12 samples seven are in desirable category with certainty level of 53-100 percent and rest of the samples belongs to acceptable category whose certainty level ranges from 67 to 96 percent. Water from these sources can be used for the drinking purpose if alternate water source is not available without any health concern on the basis of physicochemical characteristics. (author)

  7. Nuclear power plant safety

    International Nuclear Information System (INIS)

    Otway, H.J.

    1974-01-01

    Action at the international level will assume greater importance as the number of nuclear power plants increases, especially in the more densely populated parts of the world. Predictions of growth made prior to October 1973 [9] indicated that, by 1980, 14% of the electricity would be supplied by nuclear plants and by the year 2000 this figure would be about 50%. This will make the topic of international co-operation and standards of even greater importance. The IAEA has long been active in providing assistance to Member States in the siting design and operation of nuclear reactors. These activities have been pursued through advisory missions, the publication of codes of practice, guide books, technical reports and in arranging meetings to promote information exchange. During the early development of nuclear power, there was no well-established body of experience which would allow formulation of internationally acceptable safety criteria, except in a few special cases. Hence, nuclear power plant safety and reliability matters often received an ad hoc approach which necessarily entailed a lack of consistency in the criteria used and in the levels of safety required. It is clear that the continuation of an ad hoc approach to safety will prove inadequate in the context of a world-wide nuclear power industry, and the international trade which this implies. As in several other fields, the establishment of internationally acceptable safety standards and appropriate guides for use by regulatory bodies, utilities, designers and constructors, is becoming a necessity. The IAEA is presently planning the development of a comprehensive set of basic requirements for nuclear power plant safety, and the associated reliability requirements, which would be internationally acceptable, and could serve as a standard frame of reference for nuclear plant safety and reliability analyses

  8. Frequency domain reflectometry modeling for nondestructive evaluation of nuclear power plant cables

    Science.gov (United States)

    Glass, S. W.; Fifield, L. S.; Jones, A. M.; Hartman, T. S.

    2018-04-01

    Cable insulation polymers are among the more susceptible materials to age-related degradation within a nuclear power plant. This is recognized by both regulators and utilities, so all plants have developed cable aging management programs to detect damage before critical component failure in compliance with regulatory guidelines. Although a wide range of tools are available to evaluate cables and cable systems, cable aging management programs vary in how condition monitoring and nondestructive examinations are conducted as utilities search for the most reliable and cost-effective ways to assess cable system condition. Frequency domain reflectometry (FDR) is emerging as one valuable tool to locate and assess damaged portions of a cable system with minimal cost and only requires access in most cases to one of the cable terminal ends. Since laboratory studies to evaluate the use of FDR for inspection of aged cables can be expensive and data interpretation may be confounded by multiple factors which influence results, a model-based approach is desired to parametrically investigate the effect of insulation material damage in a controlled manner. This work describes development of a physics-based FDR model which uses finite element simulations of cable segments in conjunction with cascaded circuit element simulations to efficiently study a cable system. One or more segments of the cable system model have altered physical or electrical properties which represent the degree of damage and the location of the damage in the system. This circuit model is then subjected to a simulated FDR examination. The modeling approach is verified using several experimental cases and by comparing it to a commercial simulator suitable for simulation of some cable configurations. The model is used to examine a broad range of parameters including defect length, defect profile, degree of degradation, number and location of defects, FDR bandwidth, and addition of impedance-matched extensions to

  9. Evaluation of nuclear power plant siting by probabilistic assessment of environmental impacts

    International Nuclear Information System (INIS)

    Vuori, S.

    1978-01-01

    The work consists of the description of a probabilistic consequence assessment model ARANO, and the individual calculation schemes therein included. This assessment model has been applied to the risk/benefit and cost/benefit analyses of the siting of nuclear power plants. In addition, there have been made some comparisons with the alternative fossil fuelled energy production scenarios. (author)

  10. Evaluation of fatigue damage in nuclear power plants: evolution and new tools of analysis

    International Nuclear Information System (INIS)

    Cicero, R.; Corchon, F.

    2011-01-01

    This paper presents new fatigue mechanisms requiring analysis, tools developed for evaluation and the latest trends and studies that are currently working in the nuclear field, and allow proper management referring facilities the said degradation mechanism.

  11. Evaluation of environmental data relating to selected nuclear power plant sites: the Three Mile Island Nuclear Station Site

    International Nuclear Information System (INIS)

    Murarka, I.P.

    1976-08-01

    Environmental monitoring data for the years 1973 and 1974 pertaining to the Three Mile Island Nuclear Station Unit 1, which began operation in early 1974, were analyzed by the most practical qualitative and quantitative methods. Terrestrial biotic resources were considered for this plant. The effects of the operation of Unit 1 on the local terrestrial organisms were found to be undetectable. Although the plant has not operated long enough to reveal long-term deleterious effects, the present indications do not lead to a concerned prediction that any are developing. The data acquired, method of analysis, and results obtained are presented in detail along with recommendations for improving monitoring techniques

  12. The development and evaluation of programmatic performance indicators associated with maintenance at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Wreathall, J.; Fragola, J.; Appignani, P.; Burlile, G.; Shen, Y. (Science Applications International Corp., Columbus, OH (USA))

    1990-05-01

    This report summarizes the development and evaluation of programmatic performance indicators of maintenance. These indicators were selected by: (1) creating a formal framework of plant processes; (2) identifying features of plant behavior considered important to safety; (3) evaluating existing indicators against these features; and (4) performing statistical analyses for the selected indicators. The report recommends additional testing. 32 refs., 29 figs., 11 tabs.

  13. The development and evaluation of programmatic performance indicators associated with maintenance at nuclear power plants

    International Nuclear Information System (INIS)

    Wreathall, J.; Fragola, J.; Appignani, P.; Burlile, G.; Shen, Y.

    1990-05-01

    This report summarizes the development and evaluation of programmatic performance indicators of maintenance. These indicators were selected by: (1) creating a formal framework of plant processes; (2) identifying features of plant behavior considered important to safety; (3) evaluating existing indicators against these features; and (4) performing statistical analyses for the selected indicators. The report recommends additional testing. 32 refs., 29 figs., 11 tabs

  14. Installation and evaluation of a nuclear power plant Operator Advisor based on artificial intelligence technology

    International Nuclear Information System (INIS)

    Hajek, B.K.; Miller, D.W.

    1993-02-01

    The Artificial Intelligence Group in the Nuclear Engineering Program has designed and built an Operator Advisor (OA), an AI system to monitor nuclear power plant parameters, detect component and system malfunctions, dispose their causes, and provide the plant operators with the correct procedures for mitigating the consequences of the malfunctions. It then monitors performance of the procedures, and provides backup steps when specific operator actions fail. The OA has been implemented on Sun 4 workstations in Common Lisp, and has been interfaced to run in real time on the Perry Nuclear Power Plant full-function simulator in the plant training department. The eventual goal for a fully functioning Operator Advisor would be to have reactor operators receive direction for all plant operations. Such a goal requires considerable testing of the system within limited malfunction boundaries, an extensive Verification ampersand Validation (V ampersand V) effort, a large knowledge base development effort, and development of tools as part of the system to automate its maintenance. Clearly, these efforts are beyond the scope of the feasibility effort expended during this project period. However, as a result of this project, we have an AI based platform upon which a complete system can be built

  15. Methodologies for rapid evaluation of seismic demand levels in nuclear power plant structures

    International Nuclear Information System (INIS)

    Manrique, M.; Asfura, A.; Mukhim, G.

    1990-01-01

    A methodology for rapid assessment of both acceleration spectral peak and 'zero period acceleration' (ZPA) values for virtually any major structure in a nuclear power plant is presented. The methodology is based on spectral peak and ZPA amplification factors, developed from regression analyses of an analytical database. The developed amplification factors are applied to the plant's design ground spectrum to obtain amplified response parameters. A practical application of the methodology is presented. This paper also presents a methodology for calculating acceleration response spectrum curves at any number of desired damping ratios directly from a single known damping ratio spectrum. The methodology presented is particularly useful and directly applicable to older vintage nuclear power plant facilities (i.e. such as those affected by USI A-46). The methodology is based on principles of random vibration theory. The methodology has been implemented in a computer program (SPECGEN). SPECGEN results are compared with results obtained from time history analyses. (orig.)

  16. Integrated-plant-safety assessment Systematic Evaluation program. Millstone Nuclear Power Station, Unit 1, Northeast Nuclear Energy Company, Docket No. 50-245

    International Nuclear Information System (INIS)

    1982-11-01

    The Systematic Evaluation Program was initiated in February 1977 to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the Millstone Nuclear Power Station, Unit 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit 1, is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

  17. Principle evaluation on the aspects of requirements for senior operator tasks in nuclear power plants (short study)

    International Nuclear Information System (INIS)

    Birkhofer, A.; Mansfeld, G.

    1981-11-01

    The 'Guideline Relating to the Proof of the Technical Qualification of Nuclear Power Plant Personnel' specifies the general requirements for the responsible personnel at nuclear power plants. In the revised edition of this guide, issued May 1979, an engineer degree is required as a prerequisite for the senior operator qualification. The present study tries to give an objective evaluation of the requirements concerning the education and the capability level of senior operators, to cope with normal and abnormal operation conditions, with special plant conditions as well as with malfunctions or accidents. The essential arguments in the discussion on the qualification of senior operators and several alternatives how to appoint a non-engineer to the post of a senior operator are dealt with and evaluated. In doing so, psychologic, sociologic and economic aspects are taken into consideration. (orig.) [de

  18. Nuclear power plant decommissioning

    International Nuclear Information System (INIS)

    Yaziz Yunus

    1986-01-01

    A number of issues have to be taken into account before the introduction of any nuclear power plant in any country. These issues include reactor safety (site and operational), waste disposal and, lastly, the decommissioning of the reactor inself. Because of the radioactive nature of the components, nuclear power plants require a different approach to decommission compared to other plants. Until recently, issues on reactor safety and waste disposal were the main topics discussed. As for reactor decommissioning, the debates have been academic until now. Although reactors have operated for 25 years, decommissioning of retired reactors has simply not been fully planned. But the Shippingport Atomic Power Plant in Pennysylvania, the first large scale power reactor to be retired, is now being decommissioned. The work has rekindled the debate in the light of reality. Outside the United States, decommissioning is also being confronted on a new plane. (author)

  19. Nuclear power plant exports

    International Nuclear Information System (INIS)

    Degot, D.

    1987-01-01

    Framatome export expertise is discussed. Framatome can accept different types of contracts - for the supply of nuclear steam supply systems or for nuclear islands (both of which can be produced solely by Framatome) or for complete plants. Cooperation with the local industry is possible -this can involve technology transfer. Examples of exports in the following countries are given; Belgium (where there is close cooperation between Framatome and two major companies), South Africa (where Framatome has been involved in the building of a two-992 MWe unit at Koeberg), South Korea (where Framatome is building two-900 MWe nuclear islands in cooperation with Korean industry, Knu 9 and Knu 10) and China (where Framatome is to build two-1000 MWe class pressurized water reactors at Daya Bay). As well as supplying the French domestic nuclear market Framatome is a major force in the future of nuclear power in the world. (UK)

  20. Development of nuclear power plant operator simulator for man-machine interface evaluation

    International Nuclear Information System (INIS)

    Nakagawa, Takashi; Nakatani, Yoshio; Sasaki, Kazunori; Yoshikawa, Hidekazu; Takahashi, Makoto; Furuta, Tomihiko; Hasegawa, Akira.

    1997-01-01

    The operational safety in nuclear power plants depends strongly on man machine interfaces (MMI), such as assignment of equipment on control boards and operation procedures in emergency situations. Therefore, the evaluation and analysis methods for the MMI are important. In order for the methods to be practical, the methods should be executed in each step of design and be easy for designers to use. We aim to develop SEAMAID system: a computer supported system for evaluating and analyzing the MMI by simulating the interaction between the operator and the machine. In this paper, we discuss problems of the conventional methods and the required functions of the operator simulator for the SEAMAID. The operator simulator executes not human errors but correct behavior which follows the operational procedure. The SEAMAID evaluates the MMI by finding potential human errors which could occur in the simulated interactions and points out the problematic interaction parts which could induce human errors. We construct the operator simulator by combining the human model which was proposed by Prof. Reason, and the knowledge base model based on the Petri net model. This simulator can treat frequency parameter which represents the degree of frequency of using a certain knowledge. We conduct two sample simulations in different frequency parameters in the same scenario. These simulation results show that even if the operator behaves correctly following the procedure, there are alternative task sequences. We verify that the simulated interactions are in good agreement with the actual one. Also, we propose the method to point out the problematic parts in the interactions based on the working memory consumption. (author)

  1. Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Ebisawa, K.; Abe, K.; Muramatsu, K.; Itoh, M.; Kohno, K.; Tanaka, T.

    1994-01-01

    This paper presents a method for evaluating 'response factors' of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design resonse to actual response. This method has the following characteristic features: (1) The components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components. This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups. (orig.)

  2. Evaluation of hazards from industrial activity near nuclear power plants. Study of typical accidents

    International Nuclear Information System (INIS)

    Lannoy, A.; Gobert, T.; Granier, J.P.

    1981-08-01

    The design and dimensioning of nuclear power plant structures necessitate the evaluation of risks due to industrial activity. Among these risks, those due to the storage or transport of dangerous products merit special attention. They result, inter alia, in the explosion of flammable gas clouds. Such clouds can drift before igniting and, once alight, the resulting pressure wave can cause serious damage, even at a distance. A methodology both deterministic and probabilistic enabling this risk to be quantified has therefore been developed. It is based in part on an analysis of the statistics of actual accidents that have occurred. After briefly recalling the probabilistic model, the typical accidents selected are described and for three usual cases (storage under pressure, rail tank cars and road units) the main characteristics of the rupture are explicited. The deterministic models that have been worked out to calculate the consequences of such an accident: flow rate at the bursting point, evaporation, drift and atmospheric dispersion of the cloud formed, explosion of this cloud, are then described. At the present time the overpressure wave is quantified against a TNT equivalent of the explosive mixture. Some data are given as examples for three commonly employed hydrocarbons (butane, propane, propylene) [fr

  3. History of nuclear power plants safety in France (1945-2000) - Engineer techniques, expert evaluation, topical issue

    International Nuclear Information System (INIS)

    Foasso, Cyrille

    2003-01-01

    This doctoral dissertation relates the history of the mastery of risks in civil nuclear plants in France. Since 1960, it's known as the 'surete nucleaire'. Over a fifty-year period separating the discovery or the atomic fission and its industrial application on a large scale this PhD shows which technical means were used over the years by engineers to handle this risk which is said to be huge. It also studies the various processes in expert evaluation and in decision making elaborated to evaluate if the risk was acceptable or not. Beyond the conflicts between nuclear advocates and opponents, this thesis shows how ever among nuclear engineer the growing distinction between roles (promoters, experts and controlling authorities) and the various jobs (designers, builders and plant operators) triggered different estimations as far as the methods to obtain a satisfactory safety. Thanks to the progress of knowledge through research programs, thanks to the lessons drawn from the functioning or dysfunction of nuclear plants, thanks to the reinforcement of regulations (which more or less reflects the public's opinion concerning this industry) the safety has progressively improved. Thus, this historical study is multiple: a technical history of technology, a history of scientific, industrial and administrative organization, a social history and finally an international and comparative history since the nuclear energy history quickly developed beyond national boundaries. (author) [fr

  4. Nuclear Power Plant Lifetime Management Study (I)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

  5. Nuclear power plant personnel training and its evaluation. A guidebook. Executive summary

    International Nuclear Information System (INIS)

    1997-05-01

    The Guidebook will prove especially useful for, and is addressed primarily to: nuclear power operating organizations establishing or upgrading their NPP personnel training systems; regulatory personnel responsible for setting requirements and/or evaluating NPP personnel training; and organizations (within or outside the operating organization) responsible for the development, implementation and evaluation of training programmes for NPP personnel. Figs

  6. Considerations in the evaluation of concrete structures for continued service in aged Nuclear Power Plants (NPPs)

    International Nuclear Information System (INIS)

    Naus, D.; Marchbanks, M.; Oland, B.; Arndt, G.; Brown, T.

    1989-01-01

    Currently, there are /approximately/119 commercial nuclear power plants (NPPs) in the US either under construction, operating at low-to-full power, or awaiting an operating license. Together, these units have a net generating capacity of /approximately/110 GW(e). Assuming no life extension of present facilities, the operating licenses for these plants will start to expire in the middle of the next decade with Yankee Rowe being the first plant to attain this status. Where it is noted that with no life extension of facilities, a potential loss of electrical generating capacity in excess of 75 GW(e) could occur during the time period 2006 to 2020 when the operating licenses of 80 to 90 NPPs are scheduled to expire. A potential timely and cost-effective solution to meeting future electricity demand, which has worked well for non-nuclear generating plants, is to extend the service life (operating licenses) of existing NPPs. Since the concrete components in these plants provide a vital safety function, any continued service considerations must include an in-depth assessment of the safety-related concrete structures. 7 refs

  7. Criteria of diversity evaluation for intelligent diagnosis of nuclear power plants

    International Nuclear Information System (INIS)

    Washio, Takashi; Sakuma, Masatake; Furukawa, Hiroshi; Kitamura, Masaharu.

    1995-01-01

    One of important problems of a current operation support system for a nuclear power plant is that the credibility of its resultant suggestions is not always high sufficiently. The authors have proposed an efficient remedy called 'Diversity Criteria' for this issue in the previous works. It employs a variety of information resources and reasoning mechanisms for the system to enhance its entire credibility. Within this framework, a complementary combination of the resources and mechanisms is desired. The work presented here proposes systematic and quantitative measures determining the appropriate combinations. First, concrete and systematic guidelines are proposed for the detailed criteria of 'Information Diversity' and 'Methodology Diversity'. Next, two concepts of 'Orthogonality of Identified Result' and 'Orthogonality of Utilized Symptom' are presented together with their quantitative measures. These guidelines and measures have been applied to an example of failure diagnosis of a nuclear power plant, and their efficiency has been clearly confirmed. (author)

  8. Evaluation of fault coverage for digitalized system in nuclear power plants using VHDL

    International Nuclear Information System (INIS)

    Kim, Suk Joon; Lee, Jun Suk; Seong, Poong Hyun

    2003-01-01

    Fault coverage of digital systems is found to be one of the most important factors in the safety analysis of nuclear power plants. Several axiomatic models for the estimation of fault coverage of digital systems have been proposed, but to apply those axiomatic models to real digital systems, parameters that the axiomatic models require should be approximated using analytic methods, empirical methods or expert opinions. In this paper, we apply the fault injection method to VHDL computer simulation model of a real digital system which provides the protection function to nuclear power plants, for the approximation of fault detection coverage of the digital system. As a result, the fault detection coverage of the digital system could be obtained

  9. Nuclear plant license renewal

    International Nuclear Information System (INIS)

    Gazda, P.A.; Bhatt, P.C.

    1991-01-01

    During the next 10 years, nuclear plant license renewal is expected to become a significant issue. Recent Electric Power Research Institute (EPRI) studies have shown license renewal to be technically and economically feasible. Filing an application for license renewal with the Nuclear Regulatory Commission (NRC) entails verifying that the systems, structures, and components essential for safety will continue to perform their safety functions throughout the license renewal period. This paper discusses the current proposed requirements for this verification and the current industry knowledge regarding age-related degradation of structures. Elements of a license renewal program incorporating NRC requirements and industry knowledge including a schedule are presented. Degradation mechanisms for structural components, their significance to nuclear plant structures, and industry-suggested age-related degradation management options are also reviewed

  10. Evaluation of personality features of nuclear power plant operators: investigation with Minnesota Multiphasic Personality Inventory

    International Nuclear Information System (INIS)

    Sun Yiling; Liu Yulong; Li Yuan; Bian Huahui; Bi Jinling; Qiu Mengyue; Liu Chunfeng

    2011-01-01

    Objective: To explore the personality features of nuclear power plant operators and the influencing factors thereof. Methods: Minnesota Multiphasic Personality Inventory was used to examine the personality features of 136 nuclear power plant operators randomly selected from 2 cooperative units, all males. The results were compared with the nationwide norms and subsequently an inter-block contrast analysis was carried out. Results: Obvious difference was observed in the final scores between the nuclear power plant operators and nationwide norms. The former got higher scores on hysteria (t=3.05, P<0.05), and lower scores on hypochondriasis, depression, morbid personality, masculinity-femininity, paranoia, psychasthenia, schizophrenia, hypomania, and social introversion(t=7.47, 7.47, 7.31, 2.23, 15.09, 16.15, 19.28, 7.88, 11.10, P<0.05). The scores on hypochondriasis, depression, schizophrenia, and social introversion of those with the length of services over 3 years were all significantly higher than those of with the length of services less than 3 years (t=3.25, 2.51, 2.76, 3.00, P<0.05). The scores on hypochondriasis, depression, psychopathicdeviate, and social introversion of the operators aged over 30 were all significantly higher than those of the operators aged below 30 (t=2.36, 2.35, 2.01, 2.54, P<0.05). Conclusions: The psychological quality of the nuclear power plant operators is superior to that of the general population. (authors)

  11. Evaluation of the electric power production cost growth due to decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Basso, G.

    1982-01-01

    The increase of production cost for electric power generated by nuclear plants, due to their decommissioning and the end of operating life, is analysed in respect to (a) waiting time from indefinite shut-down date to the start of dismantlement, (b) financing method, (c) interest and inflation rates. The analysis shows that the additional cost is always small for those solutions which have higher probability to be adopted

  12. Development of Wall-Thinning Evaluation Procedure for Nuclear Power Plant Piping—Part 1: Quantification of Thickness Measurement Deviation

    Directory of Open Access Journals (Sweden)

    Hun Yun

    2016-06-01

    Full Text Available Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs. Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

  13. Integrated plant safety assessment systematic evaluation program. R.E. Ginna Nuclear Power Plant, Rochester Gas and Electric Corporation, Docket No. 50-244

    International Nuclear Information System (INIS)

    1982-05-01

    The Systematic Evaluation Program was initiated in February 1978 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the R.E. Ginna Nuclear Power Plant (located in Wayne County near Rochester, NY), one of ten plants reviewed under Phase II of this program, and indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

  14. Integrated plant safety assessment: systematic evaluation program. Oyster Creek nuclear generating station. GPU Nuclear Corporation and Jersey Central Power and Light Company. Docket No. 50-219

    International Nuclear Information System (INIS)

    1982-09-01

    The Systematic Evaluation Program was initiated in February 1978 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the Oyster Creek Nuclear Generating Station (located in Ocean County, New Jersey), one of ten plants reviewed under Phase II of this program, and indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

  15. Nuclear power plant of Fessenheim: evaluation of the seismic risk; Centrale Nucleaire de Fessenheim: appreciation du risque sismique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    The seismic risk taken into account during the sizing of the nuclear power plant of Fessenheim seems to have been under evaluated at this time. The revaluation of the seismic risk, as proposed, until this day by EDF in order to the third ten-year visit of the power plant, planned for 2009, leads to a significant under evaluation of the risk and then is not acceptable. The present expertise details point by point the weaknesses of these revaluation. The power plant has been sized in an elastic manner that is generally strongly for the safety side. It is imperative to proceed the most quickly as possible to a deep control of the seismic resistance of the power plant of Fessenheim and then after having proceeded to a revision of the seismic risk in taking into account the actual knowledge in this field. (N.C.)

  16. Study of a simplified method of evaluating the economic maintenance importance of components in nuclear power plant system

    International Nuclear Information System (INIS)

    Aoki, Takayuki; Takagi, Toshiyuki; Kodama, Noriko

    2014-01-01

    Safety risk importance of components in nuclear power plants has been evaluated based on the probabilistic risk assessment and used for the decisions in various plant managements. But economic risk importance of the components has not been discussed very much. Therefore, this paper discusses risk importance of the components from the viewpoint of plant economic efficiency and proposes a simplified evaluation method of the economic risk importance (or economic maintenance importance). As a result of consideration, the followings were obtained. (1) A unit cost of power generation is selected as a performance indicator and can be related to a failure rate of components in nuclear power plant which is a result of maintenance. (2) The economic maintenance importance has to major factors, i.e. repair cost at component failure and production loss associated with plant outage due to component failure. (3) The developed method enables easy understanding of economic impacts of plant shutdown or power reduction due to component failures on the plane which adopts the repair cost in vertical axis and the production loss in horizontal axis. (author)

  17. Results of the safety evaluation for the AVR-modification into a nuclear process heat plant

    International Nuclear Information System (INIS)

    Kirch, N.

    1985-01-01

    In 1983 the Juelich Nuclear Research Center (KFA) proposed the modification of the AVR for high-temperature process heat systems demonstration. This would represent the achievement of an important HTR target. The work for the modification performed so far has given evidence that the plant will continue to run reliably and has led to an optimized plant concept. Most of the investigations were devoted to safety issues. The safety and licensing questions were discussed by an advisory group of the German Federal Ministry of the Interior which gave its vote in March 1985 and came to very positive conclusions. The AVR fulfils the current safety and licensing requirements; for the proposed plant modification no severe backfitting has to be taken into account. The AVR-building and the reactor itself turned out to be earthquake-proof, even according to current licensing demands if realistic site-specific earthquake spectra are applied. Risk assessment of an airplane crash show that the public risk is negligible even in the case of unrealistically pessimistic assumptions concerning the release of radioactivity. The modified plant will have a confinement similar to the modern German HTR-design. The investigations have shown that the safety questions related to a steam reformer in a primary circuit system are solved. All consequences of process gas release into the safety enclosure or into the primary system are controlled effectively by active and passive measures. Process gas release in the vicinity of the nuclear plant is excluded by the plant concept. Furthermore, even the hypothetical assumption of process gas explosions cannot damage the essential safety functions. (author)

  18. Non-destructive electrochemical techniques applied to the corrosion evaluation of the liner structures in nuclear power plants

    International Nuclear Information System (INIS)

    Martinez, I.; Castillo, A.; Andrade, C.

    2008-01-01

    The liner structure in nuclear power plants provides containment for the operation and therefore the study of its durability and integrity during its service life is an important issue. There are several causes for the deterioration of the liner, which in general involve corrosion due to its metallic nature. The present paper is aimed at describing the assessment of corrosion problems of two liners from two different nuclear power plants, which were evaluated using non-destructive electrochemical techniques. In spite of the testing difficulties arisen, from the results extracted it can be concluded that the electrochemical techniques applied are adequate for the corrosion evaluation. They provide important information about the integrity of the structure and allow for its evolution with time to be assessed

  19. Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    Booker, C.P.; Turner, M.R.; Spore, J.W.

    1986-01-01

    The Nuclear Plant Data Bank (NPDB) is being developed at the Los Alamos National Laboratory to assist analysts in the rapid and accurate creation of input decks for reactor transient analysis. The NPDB will reduce the time and cost of the creation or modification of a typical input deck. This data bank will be an invaluable tool in the timely investigation of recent and ongoing nuclear reactor safety analysis. This paper discusses the status and plans for the NPDB development and describes its anticipated structure and capabilities

  20. Inspection-evaluation interphase in the design and construction of nuclear power plants

    International Nuclear Information System (INIS)

    Santoma, L.

    1979-01-01

    Some questions to be considered in the planning of inspection-evaluation interface are analysed. Some characteristics of a nuclear project are also analysed, such as its importance relative to safety or their frequency. The types of reviews and controls used by the various participants in a project are studied. Finally, the types of tests to be performed by regulatory agencies are classified. (author)

  1. Method of levelized discounted costs applied in economic evaluation of nuclear power plant project

    International Nuclear Information System (INIS)

    Tian Li; Wang Yongqing; Liu Jingquan; Guo Jilin; Liu Wei

    2000-01-01

    The main methods of economic evaluation of bid which are in common use are introduced. The characteristics of levelized discounted cost method and its application are presented. The method of levelized discounted cost is applied to the cost calculation of a 200 MW nuclear heating reactor economic evaluation. The results indicate that the method of levelized discounted costs is simple, feasible and which is considered most suitable for the economic evaluation of various case. The method is suggested which is used in the national economic evaluation

  2. Ardennes nuclear power plant

    International Nuclear Information System (INIS)

    1974-12-01

    The SENA nuclear power plant continued to operate, as before, at authorized rated power, namely 905MWth during the first half year and 950MWth during the second half year. Net energy production:2028GWh; hours phased to the line: 7534H; availability factor: 84%; utilization factor: 84%; total shutdowns:19; number of scrams:10; cost per KWh: 4,35 French centimes. Overall, the plant is performing very satisfactory. Over the last three years net production has been 5900GWh, corresponding to in average utilization factor of 83%

  3. Atucha I nuclear power plant azimuthal ex-vessel flux profile evaluation

    International Nuclear Information System (INIS)

    Ferraro, Diego

    2008-01-01

    Irradiation damage in RPV (Reactor Pressure Vessel) in nuclear power plants is a key parameter to be analyzed in order to assess the plant integrity up to end of life and planning for a possible plant life extension. In this work a neutronic model in MCNP that represents a sector of 30 degrees of the Atucha I power plant nucleus has been consolidated with the results of an ex-vessel dosimetry made in the outer surface of the RPV s power plant in order to analyse the irradiation damage through the dpa rate. A strong dependents of the maximum point of damage with the loading of a peripheral channel was found, so a mitigation strategy was proposed, which is basically to empty this channel and its analogs in the rest of the nucleus. Analysing this second case a notable decrease of the damage is found in the zone considerated on the model (shown through the drop of de dpa rate in the zone). [es

  4. 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.)

  5. Evaluation on applicability of new cleaning method with Ozone-solution to the laundry facility in nuclear power plant

    International Nuclear Information System (INIS)

    Tawaki, Shinji; Maeda, Toshihiko

    2000-01-01

    In nuclear power plants, the working clothes worn at the controlled area are laundered with dry cleaning and reused. Because of total abolition of chlorofluorocarbon (CFC), we have to change dry cleaning to water-cleaning. But the former inferior to the later on effectiveness of cleaning, deodorization and sterilization, and cause generation of waste fluid. And we developed a new cleaning method with Ozon-solution to solve these problem, evaluated applicability of new cleaning method to laundry facility in the plants. (author)

  6. A Study on the Measurement of Ultrasound Velocity to Evaluate Degradation of Low Voltage Cables for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Kyung Cho; Kang, Suk Chull; Goo, Cheol Soo; Kim, Jin Ho; Park, Jae Seok; Joo, Geum Jong; Park, Chi Seung

    2004-01-01

    Several kinds of low voltage cables have been used in nuclear power plants for the supply of electric power, supervision, and the propagation of control signals. These low voltage tables must be inspected for safe and stable operation of nuclear power plants. In particular, the degradation diagnosis to estimate the integrity of low voltage rabies has recently been emphasized according to the long use of nuclear power plants. In order to evaluate their degradation, the surrounding temperature, hardness of insulation material, elongation at breaking point (EAB), etc. have been used. However, the measurement of temperature or hardness is not useful because of the absence of quantitative criteria; the inspection of a sample requires turning off of the power plant power; and, the electrical inspection method is not sufficiently sensitive from the initial through the middle stage of degradation. In this research, based on the theory that the ultrasonic velocity changes with relation to the degradation of the material, we measured the ultrasonic velocity as low voltage cables were degraded. To this end, an ultrasonic degradation diagnosis device was developed and used to measure the ultrasonic velocity with the clothing on the cable, and it was confirmed that the ultrasonic velocity changes according to the degradation of low voltage cables. The low voltage cables used in nuclear power plants were degraded at an accelerated rate, and EAB was measured in a tensile test conducted after the measurement of ultrasonic velocity. With the increasing degradation degree, the ultrasonic velocity decreased, whose potential as a useful parameter for the quantitative degradation evaluation was thus confirmed

  7. A Study on the Measurement of Ultrasound Velocity to Evaluate Degradation of Low Voltage Cables for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Cho; Kang, Suk Chull; Goo, Cheol Soo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Jin Ho; Park, Jae Seok; Joo, Geum Jong; Park, Chi Seung [KAITEC, Seoul (Korea, Republic of)

    2004-08-15

    Several kinds of low voltage cables have been used in nuclear power plants for the supply of electric power, supervision, and the propagation of control signals. These low voltage tables must be inspected for safe and stable operation of nuclear power plants. In particular, the degradation diagnosis to estimate the integrity of low voltage rabies has recently been emphasized according to the long use of nuclear power plants. In order to evaluate their degradation, the surrounding temperature, hardness of insulation material, elongation at breaking point (EAB), etc. have been used. However, the measurement of temperature or hardness is not useful because of the absence of quantitative criteria; the inspection of a sample requires turning off of the power plant power; and, the electrical inspection method is not sufficiently sensitive from the initial through the middle stage of degradation. In this research, based on the theory that the ultrasonic velocity changes with relation to the degradation of the material, we measured the ultrasonic velocity as low voltage cables were degraded. To this end, an ultrasonic degradation diagnosis device was developed and used to measure the ultrasonic velocity with the clothing on the cable, and it was confirmed that the ultrasonic velocity changes according to the degradation of low voltage cables. The low voltage cables used in nuclear power plants were degraded at an accelerated rate, and EAB was measured in a tensile test conducted after the measurement of ultrasonic velocity. With the increasing degradation degree, the ultrasonic velocity decreased, whose potential as a useful parameter for the quantitative degradation evaluation was thus confirmed

  8. Systematic preparation, execution and evaluation of emergency exercises at the Beznau nuclear power plant

    International Nuclear Information System (INIS)

    Tenschert, J.

    2011-01-01

    Based on federal acts and a specific guideline of the nuclear authority ENSI, strict requirements are imposed on emergency exercises at Swiss NPPs. The Beznau NPP has conducted emergency exercises for more than 30 years. Systematic exercise planning assures that all emergency cases defined in the plant-specific emergency plan are considered in the exercise scenarios. Technically oriented scenarios cover all groups of initiating events and all safety levels of the defense in depth principle. The exercise results are an important input for optimization measures in the areas of emergency organization, documentation and infrastructure. Due to the goal-oriented enforcement of laws and guidelines by the nuclear authority ENSI, emergency exercises serve as a motor of further optimization of emergency preparedness. (orig.)

  9. Evaluation of external hazards to nuclear power plants in the United States: Other external events

    International Nuclear Information System (INIS)

    Kimura, C.Y.; Prassinos, P.G.

    1989-02-01

    In support of implementation of the Nuclear Regulatory Commission's Severe Accident Policy, the Lawrence Livermore National Laboratory (LLNL) has performed a study of the risk of core damage to nuclear power plants in the United States due to ''other external events.'' The broad objective has been to gain an understanding of whether ''other external events'' (the hazards not covered by previous reports) are among the major potential accident initiators that may pose a threat of severe reactor core damage or of large radioactive release to the environment from the reactor. The ''other external events'' covered in this report are nearby industrial/military facility accidents, on site hazardous material storage accidents, severe temperature transients, severe weather storms, lightning strikes, external fires, extraterrestrial activity, volcanic activity, earth movement, and abrasive windstorms. The analysis was based on two figures-of-merit, one based on core damage frequency and the other based on the frequency of large radioactive releases. 37 refs., 8 tabs

  10. Seismic evaluation of existing nuclear power plants and other facilities V. 2. Proceedings of the technical committee meeting. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    The objectives of this TCM are: to review the IAEA Safety Report on Seismic Evaluation of Existing Nuclear Power Plants in order to achieve a consensus among Member States on this matter and to discuss the outlines of an IAEA Co-ordinated Research Programme on specific topics related to this subject. This volume includes presentations of the member states describing the practical approach to evaluation of seismic equipment of the existing NPPs, validation of innovative systems for earthquake protection; seismic re-evaluation of the NPPs, seismic regulations and safety standards; and other activities related to seismic safety in Member States

  11. Nuclear power plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1976-01-01

    The invention concerns a quick-acting valve on the main-steam pipe of a nuclear power plant. The engineering design of the valve is to be improved. To the main valve disc, a piston-operated auxiliary valve disc is to be assigned closing a section of the area of the main valve disc. This way it is avoided that the drive of the main valve disc has to carry out different movements. 15 sub-claims. (UWI) [de

  12. Eddy-current tests on operational evaluation of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Lopez, Luiz Antonio Negro Martin; Ting, Daniel Kao Sun

    2000-01-01

    This paper presents a worldwide research on the technical and economical impacts due to failure in tube bundles of nuclear power plant steam generators. An Eddy current non destructive test using Foucault currents for the inspection and failure detection on the tubes, and also the main type of defects. The paper also presents the signals generated by a Zetec MIZ-40 test equipment. This paper also presents a brief description of an automatic system for data analysis which is under development by using a fuzzy logic and artificial intelligence

  13. Evaluation of Human Resource Needs for a New Nuclear Power Plant: Armenian Case Study

    International Nuclear Information System (INIS)

    2011-05-01

    Rising expectations of an increased role for nuclear power in providing energy for future national and global sustainable development have become a reality in many Member States of the IAEA. Over the last several years, dozens of Member States have announced plans to embark on or expand nuclear power programmes. Reflecting on these developments, the IAEA has adjusted its priorities to focus more on the nuclear power programmes of newcomers. Specifically, the IAEA has produced publications providing guidance on the development of a national infrastructure for nuclear power (IAEA Nuclear Energy Series No. NG-G-3.1) and on managing human resources in the field of nuclear energy (IAEA Nuclear Energy Series No. NG-G-2.1). Additionally, assistance to eligible Member States through new technical cooperation (TC) projects has been increased, including direct support through on-site assist visits. In 2007-2008, the IAEA carried out a TC project titled 'Feasibility study of nuclear energy development in Armenia: Evaluation of human resource needs in conjunction with new NPP build' (ARM-005). The project analysed the human resource demands required to support work at all stages of the life cycle of a new power unit planned for Armenia. This included drafting proposals for the means, conditions and requirements for development of human resource capabilities needed to carry out the work. This report is intended to complement the previous IAEA publications by providing an in-depth technical consideration into this critical area of human resource development. The report summarizes major findings of the TC project and details the tasks linked to management of the human resources that will be required by a country planning to build a new NPP. Additional guidance on the development of a national nuclear infrastructure can be found in the IAEA publication 'Milestones in the Development of a National Infrastructure for Nuclear Power', IAEA Nuclear Energy Series No. NG-G-3.1. The

  14. Evaluation of selected parameters on exposure rates in Westinghouse designed nuclear power plants

    International Nuclear Information System (INIS)

    Bergmann, C.A.

    1989-01-01

    During the past ten years, Westinghouse under EPRI contract and independently, has performed research and evaluation of plant data to define the trends of ex-core component exposure rates and the effects of various parameters on the exposure rates. The effects of the parameters were evaluated using comparative analyses or empirical techniques. This paper updates the information presented at the Fourth Bournemouth Conference and the conclusions obtained from the effects of selected parameters namely, coolant chemistry, physical changes, use of enriched boric acid, and cobalt input on plant exposure rates. The trends of exposure rates and relationship to doses is also presented. (author)

  15. Recommended criteria for the evaluation of on-site nuclear power plant emergency exercises

    International Nuclear Information System (INIS)

    Lafortune, J.F.

    1997-01-01

    A review of existing Canadian and International emergency exercise evaluation criteria and approaches has been conducted. Based on the results of the review, criteria are proposed for the evaluation of on-site emergency exercises for Canadian nuclear power stations. The proposed criteria are performance-based. They are comprehensive, yet remain adaptable to all stations and accident scenarios. They are primarily aimed at radiological emergency exercises, but are entirely applicable to fire or other conventional exercises. This report also addresses evaluation preparation and methodology. (author). 21 refs., 6 tabs

  16. 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)

  17. The Argentine nuclear policy: evaluation and proposals of the National Atomic Energy Commission. Privatization of the nuclear power plants and the situation of Atucha II. Addendum 1

    International Nuclear Information System (INIS)

    2001-01-01

    An analysis is made of the possible privatization of the nuclear power plants in Argentina. The urgent definition of a nuclear policy for the short, medium and long term is strongly recommended to the Government together with the conclusion of the construction of the Atucha II nuclear power plant

  18. Design and evaluation of augmented reality for nuclear power plant maintenance based on Cognitive activity analysis

    International Nuclear Information System (INIS)

    Koo, Jwa Jin

    2008-02-01

    The goals of maintenance in nuclear power plants (NPPs) are to reduce the number of plant shutdowns, to decrease the shutdown time, to minimize the maintenance costs, and to improve nuclear safety. Furthermore, it is difficult for maintenance personnel to conduct work in NPPs due to the complexity of NPPs, safety issues and time constraints. Human errors should be prevented to achieve maintenance goals in NPPs. Therefore, it is necessary to improve working conditions by introducing support systems. Augmented Reality (AR) has the potential to assist with maintenance tasks and make maintenance personnel concentrate on the tasks, which will then the reduced human error and improve human performance. The working procedure for AR systems was designed based on the characteristics of cognitive activity, and an experiment was conducted to estimate the effectiveness of AR. The results indicated that a well-designed working procedure for AR systems was more effective than a conventional paper-based working procedure. In addition, participants made fewer errors using the designed working procedure. AR systems as maintenance support systems are beneficial technology that can improve maintenance performance by reducing cognitive loads and unnecessary movements. An AR application can be attractive and effective at supporting maintenance task

  19. An evaluation of efforts by nuclear power plants to use ASTM D3803-89

    Energy Technology Data Exchange (ETDEWEB)

    Freeman, W.P.

    1995-02-01

    The number of nuclear power plants are now using ASTM D3803-89, {open_quotes}Standard Test Method for Nuclear-Grade Activated Carbon{close_quotes} for routine surveillance testing of adsorbents. In order to judge the impact of this change, we have gathered radioiodine removal test results from our data base on a system-by-system basis (i.e. control room, technical support center, and spent fuel pool) and compared test results obtained for the same kind of systems using the new and older test methods. Included in this comparison are systems with and without humidity control. Results are discussed from the standpoint of what to expect if a change to testing using ASTM D3803-89 is contemplated, especially regarding test results in light existing acceptance criteria. Additionally, the results are discussed from the standpoint of the sensitivity of the ASTM test method to detect when the performance of the carbon in air cleaning systems has been compromised (compared to the older methods). Finally, we offer some suggestions for how other plants might upgrade their carbon testing to incorporate testing to ASTM D3803-89.

  20. An evaluation of efforts by nuclear power plants to use ASTM D3803-89

    International Nuclear Information System (INIS)

    Freeman, W.P.

    1995-01-01

    The number of nuclear power plants are now using ASTM D3803-89, open-quotes Standard Test Method for Nuclear-Grade Activated Carbonclose quotes for routine surveillance testing of adsorbents. In order to judge the impact of this change, we have gathered radioiodine removal test results from our data base on a system-by-system basis (i.e. control room, technical support center, and spent fuel pool) and compared test results obtained for the same kind of systems using the new and older test methods. Included in this comparison are systems with and without humidity control. Results are discussed from the standpoint of what to expect if a change to testing using ASTM D3803-89 is contemplated, especially regarding test results in light existing acceptance criteria. Additionally, the results are discussed from the standpoint of the sensitivity of the ASTM test method to detect when the performance of the carbon in air cleaning systems has been compromised (compared to the older methods). Finally, we offer some suggestions for how other plants might upgrade their carbon testing to incorporate testing to ASTM D3803-89

  1. Reload safety evaluation report for Kori nuclear power plant unit 1, cycle 13

    International Nuclear Information System (INIS)

    Park, Chan Oh; Moon, Bok Ja; Cho, Byeong Ho; Nam, Kee Il; Kim, Oh Hwan; Chang, Doo Soo; Yoon, Han Young; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

    1993-03-01

    This report presents the reload safety evaluation for Kori-1, Cycle 13 and demonstrates that the reactor core being composed of various fuel assembly types applied in this evaluation will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 13 core and results are described in this report. (Author)

  2. Reload safety evaluation report for Ulchin nuclear power plant unit 1 cycle 5

    International Nuclear Information System (INIS)

    Park, Chan Oh; Kim, Yong Rae; Son, Sang Rin; Oh, Dong Seok; Kim, Hong Jin; Yoon, Kyung Ho; Yoon, Han Young; Choi, Han Rim; Choi, Dong Uk

    1992-12-01

    This report presents the reload safety evaluation for Ulchin 1, Cycle 5 and demonstrates that the reactor core being composed of various fuel assembly types applied in this evaluation will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 5 core and results are described in this report. (Author)

  3. Evaluation of radiological impact of habitual atmospheric releases in the surroundings of the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Rodriguez Rodriguez, J.M.; Rosales, I.

    1993-01-01

    The considerations and fundamental data used to evaluate the individual doses to the population living within an area of 40 km in the surroundings of Juragua nuclear power plant are presented. Recognized dispersion models (Pasquill modified) and models of dose calculations due to continuous habitual air releases composed of gases and aerosols (47 radionuclides) are used for the calculations.The following ways are taken into account for evaluation: a)irradiation of the cloud, b)irradiation of contaminated ground, c)inhalation, d) ingestion of agro-products. The iso line maps of concentrations and fundamental doses for the site are shown

  4. Risk evaluation for motor operated valves in an Inservice Testing Program at a PWR nuclear power plant in Taiwan

    International Nuclear Information System (INIS)

    Li, Y.C.; Chen, K.T.; Su, Y.L.; Ting, K.; Chien, F.T.; Li, G.D.; Huang, S.H.

    2012-01-01

    Safety related valves such as Motor Operated Valves (MOV), Air Operated Valves (AOV) or Check Valves (CV) play an important role in nuclear power plant. Functioning of these valves mainly aim at emergency reactivity control, post-accident residue heat removal, post-accident radioactivity removal and containment isolation when a design basis accident occurred. In order to maintain these valves under operable conditions, an Inservice Testing Program (IST) is defined for routine testing tasks based on the ASME Boiler and Pressure Vessel Code section XI code requirements. Risk based Inservice Testing Programs have been studied and developed extensively in the nuclear energy industry since the 1990s. Risk Based evaluations of IST can bring positive advantages to the licensee such as identifying the vulnerability of the system, reducing unnecessary testing burden, concentrating testing resources on the critical pass oriented valves and saving plant’s personnel dose exposure. This risk evaluation is incorporated with quantitative and qualitative analyses to the Motor Operated Valves under current Inservice Testing Program for PWR nuclear power plant in Taiwan. With the outcome of the risk classifications for the safety related MOVs through numerical or deterministic analyses, a risk based testing frequency relief is suggested to demonstrate the benefits received from the risk based Inservice Testing Program. The goal made of this study, it could be as a reference and cornerstone for the licensee to perform overall scope Risk-Informed Inservice Testing Program (RI-IST) evaluation by referring relevant methodologies established in this study.

  5. Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

    International Nuclear Information System (INIS)

    Widia Lastana Istanto

    2011-01-01

    Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

  6. Use of decision trees for evaluating severe accident management strategies in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclerar Engineering; Lee, Yongjin; Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of). School of Energy Systems Engineering

    2016-07-15

    Accident management strategies are defined to innovative actions taken by plant operators to prevent core damage or to maintain the sound containment integrity. Such actions minimize the chance of offsite radioactive substance leaks that lead to and intensify core damage under power plant accident conditions. Accident management extends the concept of Defense in Depth against core meltdown accidents. In pressurized water reactors, emergency operating procedures are performed to extend the core cooling time. The effectiveness of Severe Accident Management Guidance (SAMG) became an important issue. Severe accident management strategies are evaluated with a methodology utilizing the decision tree technique.

  7. Safety evaluation of the Greifswald nuclear power plant, unit 1-4

    International Nuclear Information System (INIS)

    1990-06-01

    The first interim report primarily deals with an evaluation of the pressurized components of the primary loops, especially with the embrittlement of the reactor pressure vessel material. In addition, first estimates concerning the safety design of the plants are made. The second interim report reflects the state of further studies relating to the safety design and the evaluation of operational experiences. The report includes a summarized assessment in which the recommendations cited in the technical chapters are evaluated and subdivided into three categories of backfitting measures. (orig.) [de

  8. Evaluation of the epodemiological study on children cancer in the vicinity of nuclear power plants (KiKK study). Scientific justification on the statement of the SSK (Strahlenschutzkommission)

    International Nuclear Information System (INIS)

    Grunst, Marina; Baldauf, Daniela; Reinoehl-Kompa, Sabine

    2009-01-01

    The epidemiological study on children cancer in the vicinity of nuclear power plants (KiKKK) has been en published on December 10, 2007. The study shows that there is a statistical risk for children living in the vicinity of nuclear power plants to get cancer (leukemia) before the age of 5 years. The Federal Ministry for environment, environmental protection and reactor safety has assigned the SSK (commission for radiation protection) to evaluate this study and to answer the question whether the radiation from nuclear power plants could be responsible for the observed results. The SSK statement covers the following topics: Infantile cancer, epidemiological study on cancer of children focused on leukemia; leukemia incidences in the surrounding of nuclear power plants; radiation exposure of the population and the children within the KiKKK study: report from an independent check on the recently published paper on leukemia in young children living in the vicinity of German nuclear power plants

  9. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  10. Floating nuclear power plants

    International Nuclear Information System (INIS)

    Kindt, J.W.

    1983-01-01

    This article examines the legal regime for floating nuclear power plants (FNPs), in view of the absence of specific US legislation and the very limited references to artificial islands in the Law of the Sea Convention. The environmental impacts of FNPs are examined and changes in US regulation following the Three Mile Island accident and recent US court decisions are described. References in the Law of the Sea Convention relevant to FNPs are outlined and the current status of international law on the subject is analysed. (author)

  11. Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O., E-mail: vasconv@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2016-07-01

    During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

  12. SECURE nuclear district heating plant

    International Nuclear Information System (INIS)

    Nilsson; Hannus, M.

    1978-01-01

    The role foreseen for the SECURE (Safe Environmentally Clean Urban REactor) nuclear district heating plant is to provide the baseload heating needs of primarily the larger and medium size urban centers that are outside the range of waste heat supply from conventional nuclear power stations. The rationale of the SECURE concept is that the simplicity in design and the inherent safety advantages due to the use of low temperatures and pressures should make such reactors economically feasible in much smaller unit sizes than nuclear power reactors and should make their urban location possible. It is felt that the present design should be safe enough to make urban underground location possible without restriction according to any criteria based on actual risk evaluation. From the environmental point of view, this is a municipal heat supply plant with negligible pollution. Waste heat is negligible, gaseous radioactivity release is negligible, and there is no liquid radwaste release. Economic comparisons show that the SECURE plant is competitive with current fossil-fueled alternatives. Expected future increase in energy raw material prices will lead to additional energy cost advantages to the SECURE plant

  13. Chemistry management system for nuclear power plants

    International Nuclear Information System (INIS)

    Nagasawa, Katsumi; Maeda, Katsuji

    1998-01-01

    Recently, the chemistry management in the nuclear power plants has been changing from the problem solution to the predictive diagnosis and maintenance. It is important to maintain the integrity of plant operation by an adequate chemistry control. For these reasons, many plant operation data and chemistry analysis data should be collected and treated effectively to evaluate chemistry condition of the nuclear power plants. When some indications of chemistry anomalies occur, quick and effective root cause evaluation and countermeasures should be required. The chemistry management system has been developed as to provide sophisticate chemistry management in the nuclear power plants. This paper introduces the concept and functions of the chemistry management system for the nuclear power plants. (author)

  14. Evaluation of static analysis tools used to assess software important to nuclear power plant safety

    International Nuclear Information System (INIS)

    Ourghanlian, Alain

    2015-01-01

    We describe a comparative analysis of different tools used to assess safety-critical software used in nuclear power plants. To enhance the credibility of safety assessments and to optimize safety justification costs, Electricit e de France (EDF) investigates the use of methods and tools for source code semantic analysis, to obtain indisputable evidence and help assessors focus on the most critical issues. EDF has been using the PolySpace tool for more than 10 years. Currently, new industrial tools based on the same formal approach, Abstract Interpretation, are available. Practical experimentation with these new tools shows that the precision obtained on one of our shutdown systems software packages is substantially improved. In the first part of this article, we present the analysis principles of the tools used in our experimentation. In the second part, we present the main characteristics of protection-system software, and why these characteristics are well adapted for the new analysis tools.

  15. Use of probabilistic safety assessment to evaluate nuclear power plant technical specifications

    International Nuclear Information System (INIS)

    1991-04-01

    The meeting was organized to promote the use of risk and reliability techniques to improve safety and operational flexibility of nuclear power plants. The 43 participants from 17 countries and IAEA took part in the meeting. The 23 papers which were presented indicated that there was a worldwide recognition that risk and reliability techniques could be a useful tool to optimize technical specification requirements. A separate abstract was prepared for each of these papers. This TECDOC, prepared by the meeting participants, summarizes the insights from the various papers presented and the plenary discussions. It also presents a regulatory perspective for the use of PSA-based Technical Specifications in Member States. Finally, the proposal of a pilot study programme on the use of risk and reliability methods for technical specifications optimization to be undertaken by Member States under the coordination of the IAEA was discussed. Refs, figs and tabs

  16. Development and validation of a heuristic model for evaluation of the team performance of operators in nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Sa Kil; Byun, Seong Nam; Lee, Dhong Hoon

    2011-01-01

    Highlights: → We develop an estimation model for evaluation of the team performance of MCR. → To build the model, we extract team performance factors through reviewing literatures and identifying behavior markers. → We validate that the model is adaptable to the advanced MCR of nuclear power plants. → As a result, we find that the model is a systematic and objective to measure team performance. - Abstract: The global concerns about safety in the digital technology of the main control room (MCR) are growing as domestic and foreign nuclear power plants are developed with computerized control facilities and human-system interfaces. In a narrow space, the digital technology contributes to a control room environment, which can facilitate the acquisition of all the information needed for operation. Thus, although an individual performance of the advanced MCR can be further improved; there is a limit in expecting an improvement in team performance. The team performance depends on organic coherence as a whole team rather than on the knowledge and skill of an individual operator. Moreover, a good team performance improves communication between and within teams in an efficient manner, and then it can be conducive to addressing unsafe conditions. Respecting this, it is important and necessary to develop methodological technology for the evaluation of operators' teamwork or collaboration, thus enhancing operational performance in nuclear power plant at the MCR. The objectives of this research are twofold: to develop a systematic methodology for evaluation of the team performance of MCR operators in consideration of advanced MCR characteristics, and to validate that the methodology is adaptable to the advanced MCR of nuclear power plants. In order to achieve these two objectives, first, team performance factors were extracted through literature reviews and methodological study concerning team performance theories. Second, the team performance factors were identified and

  17. Obrigheim nuclear power plant

    International Nuclear Information System (INIS)

    1976-05-01

    The gross output of the 345MWe pressurized water nuclear power station at Obrigheim, operation on base load, amounted to about 2.57TWh in 1974, the net power fed to the grid being about 2.44TWh. The core was used to its full capacity until 10 May 1974. Thereafter, the reactor was on stretch-out operation with steadily decreasing load until refuelled in August 1974. Plant availability in 1974 amounted to 92.1%. Of the 7.9% non-availability, 7.87% was attributable to the refuelling operation carried out from 16 August to 14 September and to the inspection, overhaul and repair work and the routine tests performed during this period. The plant was in good condition. Only two brief shutdowns occurred in 1974, the total outage time being 21/2 hours. From the beginning of trial operation in March 1969 to the end of 1974, the plant achieved an availability factor of 85.2%. The mean core burnup at the end of the fifth cycle was 19600 MWd/tonne U, with one fuel element that had been used for four cycles achieving a mean burnup of 39000 MWd/tonne U. The sipping test on the fuel elements revealed defective fuel-rods in a prototype plutonium fuel element, a high-efficiency uranium fuel element and a uranium fuel element. The quantities of radioactive substances released to the environment in 1974 were far below the officially permitted values. In july 1974, a reference preparation made up in the nuclear power station in October 1973 was discovered by outsiders on the Obrigheim municipality rubbish tip. The investigations revealed that this reference preparation had very probably been abstracted from the plant in October 1973 and arrived at the rubbish tip in a most irregular manner shortly before its discovery

  18. Evaluation of environmental data relating to selected nuclear power plant sites. The Zion Nuclear Power Station site

    Energy Technology Data Exchange (ETDEWEB)

    Murarka, I.P.; Policastro, A.; Daniels, E.; Ferrante, J.; Vaslow, F.

    1976-11-01

    Environmental monitoring data for the years 1972 through 1975 pertaining to the Zion Station, which began commercial operation in 1974, were analyzed by the most practical qualitative and quantitative methods. Thermal plume, water chemistry and aquatic biota data were evaluated and the results are presented. The results showed no significant immediate deleterious effects, thus confirming preoperational predictions. Although the plant has not operated long enough to reveal long-term deleterious effects, the present indications do not lead to a concerned prediction that any are developing. The data utilized, the methods of analysis, and the results obtained are presented in detail, along with recommendations for improving monitoring techniques.

  19. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Byrd, J.S.; Fisher, J.J.; DeVries, K.R.; Martin, T.P.

    1987-01-01

    Expert robots will enhance safety and operations in nuclear plants. E. I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligent, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  20. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Byrd, J.S.; Fisher, J.J.; DeVries, K.R.; Martin, T.P.

    1987-01-01

    Expert robots enhance a safety and operations in nuclear plants. E.I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligence, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  1. Appraisal of nuclear power plant reliability

    International Nuclear Information System (INIS)

    Applegren, G.H.; Beckwith, R.; Jakubowski, C.L.; Whysong, J.L.; Rossin, A.D.

    1975-01-01

    The performance of large power plants at Commonwealth Edison is evaluated. There are seven nuclear units with a total capacity of about 5500 MW(e); six of the units are 800 MW(e) in size or larger. Fossil-fired capacity totals about 10,600 MW, primarily coal-fired. There are five coal-fired units 600 MW in size or larger. All large units have been installed since 1964. It is found that the nuclear plants in the system are more economical than coal-fired plants even at capacity factors well below 65 percent; that the nuclear plants have slightly better availability than the fossil plants; and that system characteristics have a strong influence on the performance of all of the plants. A cost comparison is shown of electric power generation from nuclear, a combination of Illinois and Western coal, and Western coal plants. (U.S.)

  2. Safety evaluation report related to the operation of Sequoyah Nuclear Plant, Units 1 and 2, Docket Nos. 50-327 and 50-328, Tennessee Valley Authority

    International Nuclear Information System (INIS)

    1982-12-01

    Supplement No. 6 to the Safety Evaluation Report (SER) related to the operation of the Tennessee Valley Authority's Sequoyah Nuclear Plant, Units 1 and 2, located in Hamilton County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The purpose of this supplement is to update the staff's evaluations of the issues related to the hydrogen mitigation system identified in the SER and previous supplements as needing resolution

  3. Environmental evaluation of a nuclear power plant on Lake Erie: some aquatic impacts

    International Nuclear Information System (INIS)

    Reutter, J.M.

    1976-01-01

    The Toledo Edison Company and The Cleveland Electric Illuminating Company are currently building the first unit of the Davis--Besse Nuclear Power Station on the south shore of Lake Erie at Locust Point. This plant will utilize water from Lake Erie for cooling purposes and to replenish the cooling tower blowdown which will be returned to the lake at the maximum of 11.1 0 C above ambient. Laboratory experiments were conducted on Lake Erie fish to determine their seasonal final temperature preferenda and the effects of sudden temperature changes. These results were correlated with the existing fish community at Locust Point to predict the effects of the thermal discharge from the Davis--Besse Nuclear Power Station on the surrounding fish community and to predict the effects of thermal discharges, in general, on the Lake Erie fishery resource. Phytoplankton populations, zooplankton populations and several water quality parameters were measured at Locust Point in 1974 and 1975. Correlation coefficients and multiple regressions were computed using these plankton populations and the water quality parameters to determine the extent to which the existing plankton populations are effected by the water quality. Over 2000 fish representing 24 species were tested to determine their final temperature preferenda

  4. Reload safety evaluation report for Ulchin nuclear power plant unit 2, cycle 6

    International Nuclear Information System (INIS)

    Chung, Jin Gon; Park, Jin Ha; Kim, Oh Hwan; Oh, Dong Seok; Kim, Du Ill; Choi, Han Rim; Ku, Dong Uk; Bae, Hoo Gun

    1994-07-01

    This report presents a reload safety evaluation for Ulchin-2, cycle 6 and demonstrates that the core being composed of various fuel types as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the cycle 6 core and results are described in the report. (Author) 1 ref., 3 figs., 7 tabs

  5. Reload safety evaluation report for Ulchin nuclear power plant unit 1, cycle 7

    International Nuclear Information System (INIS)

    Kim, Yong Rae; Kwon, Hyuk Sung; Kim, Oh Hwan; Choi, Han Rim; Yoon, Han Young; Ku, Dong Uk; Suh, Jung Min; Bae, Hoo Gun

    1995-02-01

    This report presents a reload safety evaluation for UCN-2, Cycle 7 and demonstrates that the core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 7 core and results are described in this report. (Author) 1 refs., 3 figs., 6 tabs

  6. Reload safety evaluation report for yonggwang nuclear power plant unit 1 cycle 7

    International Nuclear Information System (INIS)

    Park, Chan Oh; Kwon, Tae Je; Park, Sang Yoon; Sung, Kang Sik; Kim, Ki Hang; Yim, Jeong Sik; Kim, Du Ill; Choi, Han Rim; Bae, Hoo Gun

    1992-06-01

    This report presents the reload safety evaluation for YGN-1, Cycle 7 and demonstrates that the reactor core being entirely composed of KOFA as discribed below will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 7 core and results are described in this report. (Author)

  7. Reload safety evaluation report for ulchin nuclear power plant unit 2, cycle 4

    International Nuclear Information System (INIS)

    Park, Chan Oh; Park, Yong Soo; Kim, Hong Jin; Kim, Il Kon; Oh, Dong Seok; Yoon, Han Yong; Choi, Han Rim; Choi, Dong Uk; Lee, Chung Chan; Zee, Sung Kyun

    1992-09-01

    This report presents a reload safety evaluation for Ulchin-2, Cycle 4 and demonstrates that the core being composed of various fuel types as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 4 core and results are described in the report. (Author)

  8. Reload safety evaluation report for yonggwang nuclear power plant unit 2 cycle 7

    International Nuclear Information System (INIS)

    Zee, Sung Kyun; Choi, Gyoo Hwan; Lee, Ki Bog; Park, Sang Yoon

    1993-01-01

    This report presents the reload safety evaluation for YGN-2, Cycle 7 and demonstrates that the reactor core being entirely composed of KOFA as described below will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which would potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 7 core design described herein. (Author)

  9. Reload safety evaluation report for Yonggwang nuclear power plant unit 1, cycle 8

    International Nuclear Information System (INIS)

    Lee, Won Jae; Yoon, Kyung Ho; Cho, Young Chul; Kim, Jae Hak; Um, Kil Sup; Choi, Han Rim; Kim, Ki Hang; Sung, Kang Sik

    1993-09-01

    This report presents a reload safety evaluation for YGN-1, cycle 8 and demonstrates that the core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the cycle 8 core and results are described in the report. (Author) 1 ref., 4 figs., 5 tabs

  10. Reload safety evaluation report for Ulchin nuclear power plant unit 1, cycle 6

    International Nuclear Information System (INIS)

    Lee, Won Jae; Kim, Yong Rae; Kim, Oh Hwan; Kwon, Hyuk Sung; Yoon, Han Young; Choi, Han Rim; Ku, Dong Uk

    1993-11-01

    This report presents a reload safety evaluation for Ulchin 1, cycle 6 and demonstrates that the reactor core being fully composed of KOFA as described in this report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the cycle 6 core and results are described in this report. (Author) 1 ref., 5 figs., 6 tabs

  11. Reload safety evaluation report for kori nuclear power plant unit 4, cycle 8

    International Nuclear Information System (INIS)

    Park, Chan Oh; Jung, Yil Sup; Kim, Si Yong; Kim, Ki Hang; Kwon, Hyuk Sung; Oh, Dong Seok; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

    1993-06-01

    This report presents the reload safety evaluation for Kori-4, Cycle 8 and demonstrate that the reactor core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licening bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 8 core design described herein. (Author)

  12. Reload safety evaluation report for Yonggwang nuclear power plant unit 1, cycle 9

    International Nuclear Information System (INIS)

    Cho, Young Chul; Nam, Kee Il; Kim, Ki Hang; Suh, Jung Min; Um, Kil Sup; Ban, Chang Hwan; Bae, Hoo Gun

    1995-02-01

    This report presents a reload safety evaluation for YGN-1, Cycle 9 and demonstrates that the core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 9 core and results are described in this report. (Author) 1 refs., 3 figs., 6 tabs

  13. Evaluation tests on controbloc, a programmable automaton for nuclear power plants

    International Nuclear Information System (INIS)

    Pralus, B.; Bourassin, J.L.; Varaldi, G.

    1983-01-01

    Controbloc is the programmable automaton used by Electricite de France (EDF) to equip its 1300 MW range of power plants. EDF and the designer CGEE Alsthom have conducted a large number of tests on prototype series equipment to determine whether it matches its specifications. These tests were performed in various laboratories in the Paris area and were concerned with: (1) verifying performance (acquisition, processing and retrieval cycle times - multiplexed exchange performance); (2) the robustness of equipment and its behaviour in industrial conditions (climatic tests in line with IEC regulations - earthquake resistance in particularly severe conditions - resistance to industrial interference); (3) behaviour when internal faults occur, an aspect of particular importance in view of the role played by Controbloc in nuclear power plants; in the event of a failure, safety (incorrect commands must not be given) and availability (resumption of operation) must be guaranteed. A large number of checks have been carried out both by the designer and EDF, the latter having put into operation an original method which is described in the paper. Controbloc has been thoroughly tested and its modular construction has facilitated the performance of these tests. Some shortcomings have been revealed which the designer or EDF has remedied. (author)

  14. Fuzzy inference system for evaluating and improving nuclear power plant operating performance

    International Nuclear Information System (INIS)

    Guimaraes, Antonio Cesar F.; Lapa, Celso Marcelo Franklin

    2003-01-01

    This paper presents a fuzzy inference system (FIS) as an approach to estimate Nuclear Power Plant (NPP) performance indicators. The performance indicators for this study are the energy availability factor (EAF) and the planned (PUF) and unplanned unavailability factor (UUF). These indicators are obtained from a non analytical combination among the same operational parameters. Such parameters are, for example, environment impacts, industrial safety, radiological protection, safety indicators, scram rate, thermal efficiency, and fuel reliability. This approach uses the concept of a pure fuzzy logic system where the fuzzy rule base consists of a collection of fuzzy IF-THEN rules. The fuzzy inference engine uses these fuzzy IF-THEN rules to determine a mapping from fuzzy sets in the input universe of discourse to fuzzy sets in the output universe of discourse based on fuzzy logic principles. The results demonstrated the potential of the fuzzy inference to generate a knowledge basis that correlate operations occurrences and NPP performance. The inference system became possible the development of the sensitivity studies, future operational condition previsions and may support the eventual corrections on operation of the plant

  15. Performance evaluation of drip-fertigated cotton grown under different plant densities using nuclear techniques

    International Nuclear Information System (INIS)

    Janat, M.; Kalhout, A.

    2007-04-01

    Field experiment was conducted over two growing seasons to assess different planting densities of cotton variety Aleppo 118 (71.000, 57.000, 48.000, 41.000, 33.500 plants /ha), and two irrigation systems; one irrigation line per one planting row and one irrigation line per two planting rows. Nitrogen fertilizer (120 kg N/ha) as Urea (46% N) was injected through the irrigation system in six equally split applications. A labeled area (1.0 m 2 ) was established for the labeled sub plots and labeled Urea was applied to the labeled sub plots in the same manner as for unlabeled N fertilizer. Irrigation scheduling was accomplished using the direct method of neutron scattering technique. Irrigation was determined when the soil moisture content in the active root depth reached almost 80% of the field capacity. The amount of water applied for one line / one row were 6738 and 9149 m 3 /ha, whereas, for one line/two rows were 7489 and 12653 m 3 / ha for the two growing seasons 2004 and 2005 respectively. The objectives of the experiment were to evaluate the effect of different planting densities and two irrigation system on cotton yield, lint properties, dry matter yield, N-uptake, chlorophyll content and leaf area. The experimental design was randomized block design with 6 replications for each irrigation method. Results revealed that no significant differences between all different plant densities were recorded for all growth parameters tested in this study such as seed cotton yield, dry matter yield, lint properties, chlorophyll content and leaf area.(author)

  16. Methods of assessing nuclear power plant risks

    International Nuclear Information System (INIS)

    Skvarka, P.; Kovacz, Z.

    1985-01-01

    The concept of safety evalution is based on safety criteria -standards or set qualitative values of parameters and indices used in designing nuclear power plants, incorporating demands on the quality of equipment and operation of the plant, its siting and technical means for achieving nuclear safety. The concepts are presented of basic and optimal risk values. Factors are summed up indispensable for the evaluation of the nuclear power plant risk and the present world trend of evaluation based on probability is discussed. (J.C.)

  17. Integrated Plant Safety Assessment: Systematic Evaluation Program. Yankee Nuclear Power Station, Yankee Atomic Electric Company, Docket No. 50-29. Final report

    International Nuclear Information System (INIS)

    1983-06-01

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

  18. Integrated Plant Safety Assessment, Systematic Evaluation Program. Yankee Nuclear Power Station, Yankee Atomic Electric Company, Docket No. 50-29. Draft report

    International Nuclear Information System (INIS)

    1983-02-01

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

  19. Conformation of an evaluation process for a license renovation solicitude of a nuclear power plant in Mexico. Part 2

    International Nuclear Information System (INIS)

    Serrano R, M. de L.

    2013-10-01

    At the present time the operation licenses in force for the reactors of the Nuclear Power Plant of Laguna Verde (NPP-L V) will expire in the year 2020 and 2025 for the Unit-1 and Unit-2, respectively, for which the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) has begun its preparation to assist a solicitude of the licensee to continue the operation of the NPP-L V. The present work has the purpose of defining the steps to continue and to generate the documents that would help in this process, as the normative, guides, procedures, regulations, controls, etc. so that the evaluation process will be effective and efficient, as much for the regulator organ as for the licensee. The advance carried out in the continuation of the conformation of an evaluation process of license renovation solicitude is also exposed, taking like base the requirements established by the CNSNS, the regulator organ of the United States (US NRC), and the IAEA for license renovation solicitude of this type. A summary of the licenses granted from the beginning of commercial operation of the NPP-L V is included, both units and the amendments to these licenses, explaining the reason of the amendment shortly and in the dates they were granted. A brief exposition of the nuclear power plants to world level that have received extension of its operation is included. The normative that can be applied in a life extension evaluation is presented, the evaluation process to continue with the guides of the US NRC, the reach of the evaluation and the minimum information required to the licensee that should accompany to their solicitude. (author)

  20. Status of evaluations and modifications of diesel generator status annunciator systems at various US nuclear power plants

    International Nuclear Information System (INIS)

    Shindell, B.M.; Rumble, R.P.

    1979-10-01

    This report documents the current status of evaluations and modifications to the diesel generator status annunciators in a number of US nuclear power plants. These modifications may be required in order to: ensure that all conditions which might render the diesel generators incapable of automatic starting are annunciated in the control room; ensure that the wording on the control room annunciator clearly indicates to the operator that the diesel generator is unavailable if such is the case; and separate disabling and non-disabling annunciation

  1. Computer–Based Procedures for Nuclear Power Plant Field Workers: Preliminary Results from Two Evaluation Studies

    Energy Technology Data Exchange (ETDEWEB)

    Katya L Le Blanc; Johanna H Oxstrand

    2013-10-01

    The Idaho National Laboratory and participants from the U.S. nuclear industry are collaborating on a research effort aimed to augment the existing guidance on computer-based procedure (CBP) design with specific guidance on how to design CBP user interfaces such that they support procedure execution in ways that exceed the capabilities of paper-based procedures (PBPs) without introducing new errors. Researchers are employing an iterative process where the human factors issues and interface design principles related to CBP usage are systematically addressed and evaluated in realistic settings. This paper describes the process of developing a CBP prototype and the two studies conducted to evaluate the prototype. The results indicate that CBPs may improve performance by reducing errors, but may increase the time it takes to complete procedural tasks.

  2. Evaluation of earthquake vibration on aseismic design of nuclear power plant judging from recent earthquakes

    International Nuclear Information System (INIS)

    Dan, Kazuo

    2006-01-01

    The Regulatory Guide for Aseismic Design of Nuclear Reactor Facilities was revised on 19 th September, 2006. Six factors for evaluation of earthquake vibration are considered on the basis of the recent earthquakes. They are 1) evaluation of earthquake vibration by method using fault model, 2) investigation and approval of active fault, 3) direct hit earthquake, 4) assumption of the short active fault as the hypocentral fault, 5) locality of the earthquake and the earthquake vibration and 6) remaining risk. A guiding principle of revision required new evaluation method of earthquake vibration using fault model, and evaluation of probability of earthquake vibration. The remaining risk means the facilities and people get into danger when stronger earthquake than the design occurred, accordingly, the scattering has to be considered at evaluation of earthquake vibration. The earthquake belt of Hyogo-Nanbu earthquake and strong vibration pulse in 1995, relation between length of surface earthquake fault and hypocentral fault, and distribution of seismic intensity of off Kushiro in 1993 are shown. (S.Y.)

  3. Sabotage at Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Purvis, James W.

    1999-07-21

    Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented.

  4. Sabotage at Nuclear Power Plants

    International Nuclear Information System (INIS)

    Purvis, James W.

    1999-01-01

    Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented

  5. Nuclear power plant human computer interface design incorporating console simulation, operations personnel, and formal evaluation techniques

    International Nuclear Information System (INIS)

    Chavez, C.; Edwards, R.M.; Goldberg, J.H.

    1993-01-01

    New CRT-based information displays which enhance the human machine interface are playing a very important role and are being increasingly used in control rooms since they present a higher degree of flexibility compared to conventional hardwired instrumentation. To prototype a new console configuration and information display system at the Experimental Breeder Reactor II (EBR-II), an iterative process of console simulation and evaluation involving operations personnel is being pursued. Entire panels including selector switches and information displays are simulated and driven by plant dynamical simulations with realistic responses that reproduce the actual cognitive and physical environment. Careful analysis and formal evaluation of operator interaction while using the simulated console will be conducted to determine underlying principles for effective control console design for this particular group of operation personnel. Additional iterations of design, simulation, and evaluation will then be conducted as necessary

  6. Nuclear plant analyzer desktop workstation

    International Nuclear Information System (INIS)

    Beelman, R.J.

    1990-01-01

    In 1983 the U.S. Nuclear Regulatory Commission (USNRC) commissioned the Idaho National Engineering Laboratory (INEL) to develop a Nuclear Plant Analyzer (NPA). The NPA was envisioned as a graphical aid to assist reactor safety analysts in comprehending the results of thermal-hydraulic code calculations. The development was to proceed in three distinct phases culminating in a desktop reactor safety workstation. The desktop NPA is now complete. The desktop NPA is a microcomputer based reactor transient simulation, visualization and analysis tool developed at INEL to assist an analyst in evaluating the transient behavior of nuclear power plants by means of graphic displays. The NPA desktop workstation integrates advanced reactor simulation codes with online computer graphics allowing reactor plant transient simulation and graphical presentation of results. The graphics software, written exclusively in ANSI standard C and FORTRAN 77 and implemented over the UNIX/X-windows operating environment, is modular and is designed to interface to the NRC's suite of advanced thermal-hydraulic codes to the extent allowed by that code. Currently, full, interactive, desktop NPA capabilities are realized only with RELAP5

  7. Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk Methods Integration and Evaluation Program (RMIEP)

    International Nuclear Information System (INIS)

    Payne, A.C. Jr.; Eide, S.A.; LaChance, J.C.; Whitehead, D.W.

    1992-10-01

    This volume presents the results of the initiating event and accident sequence delineation analyses of the LaSalle Unit II nuclear power plant performed as part of the Level III PRA being performed by Sandia National Laboratories for the Nuclear Regulatory Commission. The initiating event identification included a thorough review of extant data and a detailed plant specific search for special initiators. For the LaSalle analysis, the following initiating events were defined: eight general transients, ten special initiators, four LOCAs inside containment, one LOCA outside containment, and two interfacing LOCAs. Three accident sequence event trees were constructed: LOCA, transient, and ATWS. These trees were general in nature so that a tree represented all initiators of a particular type (i.e., the LOCA tree was constructed for evaluating small, medium, and large LOCAs simultaneously). The effects of the specific initiators on the systems and the different success criteria were handled by including the initiating events directly in the system fault trees. The accident sequence event trees were extended to include the evaluation of containment vulnerable sequences. These internal event accident sequence event trees were also used for the evaluation of the seismic, fire, and flood analyses

  8. Development of a method to evaluate shared alternate AC power source effects in multi-unit nuclear power plants

    International Nuclear Information System (INIS)

    Jung, Woo Sik; Yang, Joon Eun

    2003-07-01

    In order to evaluate accurately a Station BlackOut (SBO) event frequency of a multi-unit nuclear power plant that has a shared Alternate AC (AAC) power source, an approach has been developed which accommodates the complex inter-unit behavior of the shared AAC power source under multi-unit Loss Of Offsite Power (LOOP) conditions. The approach is illustrated for two cases, 2 units and 4 units at a single site, and generalized for a multi-unit site. Furthermore, the SBO frequency of the first unit of the 2-unit site is quantified. The SBO frequency at a target unit of Probabilistic Safety Assessment (PSA) could be underestimated if the inter-unit dependency of the shared AAC power source is not properly modeled. The effect of the inter-unit behavior of the shared AAC power source on the SBO frequency is not negligible depending on the Common Cause Failure (CCF) characteristics among AC power sources. The methodology suggested in the present report is believed to be very useful in evaluating the SBO frequency and the core damage frequency resulting from the SBO event. This approach is also applicable to the probabilistic evaluation of the other shared systems in a multi-unit nuclear power plant

  9. Light Water Reactor Sustainability Program: Computer-Based Procedures for Field Activities: Results from Three Evaluations at Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Le Blanc, Katya [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bly, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    The Computer-Based Procedure (CBP) research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which is a research and development (R&D) program sponsored by Department of Energy (DOE) and performed in close collaboration with industry R&D programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. One area that could yield tremendous savings in increased efficiency and safety is in improving procedure use. Nearly all activities in the nuclear power industry are guided by procedures, which today are printed and executed on paper. This paper-based procedure process has proven to ensure safety; however, there are improvements to be gained. Due to its inherent dynamic nature, a CBP provides the opportunity to incorporate context driven job aids, such as drawings, photos, and just-in-time training. Compared to the static state of paper-based procedures (PBPs), the presentation of information in CBPs can be much more flexible and tailored to the task, actual plant condition, and operation mode. The dynamic presentation of the procedure will guide the user down the path of relevant steps, thus minimizing time spent by the field worker to evaluate plant conditions and decisions related to the applicability of each step. This dynamic presentation of the procedure also minimizes the risk of conducting steps out of order and/or incorrectly assessed applicability of steps.

  10. Threatened and endangered species evaluation for 75 licensed commercial nuclear power generating plants

    Energy Technology Data Exchange (ETDEWEB)

    Sackschewsky, M.R.

    1997-03-01

    The Endangered Species Act (ESA) of 1973, as amended, and related implementing regulations of the jurisdictional federal agencies, the U.S. Departments of Commerce and Interior, at 50 CFR Part 17. 1, et seq., require that federal agencies ensure that any action authorized, funded, or carried out under their jurisdiction is not likely to jeopardize the continued existence of any threatened or endangered species or result in the destruction or adverse modification of critical habitats for such species. The issuance and maintenance of a federal license, such as a construction permit or operating license issued by the U.S. Nuclear Regulatory Commission (NRC) for a commercial nuclear power generating facility is a federal action under the jurisdiction of a federal agency, and is therefore subject to the provisions of the ESA. The U.S. Department of the Interior (through the Fish and Wildlife Service), and the U.S. Department of Commerce, share responsibility for administration of the ESA. The National Marine Fisheries Service (NMFS) deals with species that inhabit marine environments and anadromous fish, while the U.S. Fish and Wildlife Service (USFWS) is responsible for terrestrial and freshwater species and migratory birds. A species (or other distinct taxonomic unit such as subspecies, variety, and for vertebrates, distinct population units) may be classified for protection as `endangered` when it is in danger of extinction within the foreseeable future throughout all or a significant portion of its range. A `threatened` classification is provided to those animals and plants likely to become endangered within the foreseeable future throughout all or a significant portion of their ranges. As of February 1997, there were about 1067 species listed under the ESA in the United States. Additionally there were approximately 125 species currently proposed for listing as threatened or endangered, and another 183 species considered to be candidates for formal listing proposals.

  11. Threatened and endangered species evaluation for 75 licensed commercial nuclear power generating plants

    International Nuclear Information System (INIS)

    Sackschewsky, M.R.

    1997-03-01

    The Endangered Species Act (ESA) of 1973, as amended, and related implementing regulations of the jurisdictional federal agencies, the U.S. Departments of Commerce and Interior, at 50 CFR Part 17. 1, et seq., require that federal agencies ensure that any action authorized, funded, or carried out under their jurisdiction is not likely to jeopardize the continued existence of any threatened or endangered species or result in the destruction or adverse modification of critical habitats for such species. The issuance and maintenance of a federal license, such as a construction permit or operating license issued by the U.S. Nuclear Regulatory Commission (NRC) for a commercial nuclear power generating facility is a federal action under the jurisdiction of a federal agency, and is therefore subject to the provisions of the ESA. The U.S. Department of the Interior (through the Fish and Wildlife Service), and the U.S. Department of Commerce, share responsibility for administration of the ESA. The National Marine Fisheries Service (NMFS) deals with species that inhabit marine environments and anadromous fish, while the U.S. Fish and Wildlife Service (USFWS) is responsible for terrestrial and freshwater species and migratory birds. A species (or other distinct taxonomic unit such as subspecies, variety, and for vertebrates, distinct population units) may be classified for protection as 'endangered' when it is in danger of extinction within the foreseeable future throughout all or a significant portion of its range. A 'threatened' classification is provided to those animals and plants likely to become endangered within the foreseeable future throughout all or a significant portion of their ranges. As of February 1997, there were about 1067 species listed under the ESA in the United States. Additionally there were approximately 125 species currently proposed for listing as threatened or endangered, and another 183 species considered to be candidates for formal listing proposals

  12. Nuclear power plant operator licensing

    International Nuclear Information System (INIS)

    1997-01-01

    The guide applies to the nuclear power plant operator licensing procedure referred to the section 128 of the Finnish Nuclear Energy Degree. The licensing procedure applies to shift supervisors and those operators of the shift teams of nuclear power plant units who manipulate the controls of nuclear power plants systems in the main control room. The qualification requirements presented in the guide also apply to nuclear safety engineers who work in the main control room and provide support to the shift supervisors, operation engineers who are the immediate superiors of shift supervisors, heads of the operational planning units and simulator instructors. The operator licensing procedure for other nuclear facilities are decided case by case. The requirements for the basic education, work experience and the initial, refresher and complementary training of nuclear power plant operating personnel are presented in the YVL guide 1.7. (2 refs.)

  13. Computer-based procedure for field activities: Results from three evaluations at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); bly, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); LeBlanc, Katya [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

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

  14. Aging surveillance and remanent lifetime evaluation of Nuclear Power Plants components

    International Nuclear Information System (INIS)

    Aguado, M.; Cueto-Felgueroso, C.

    1997-01-01

    A lot of resources have been invested in the last years by the electric sector, in order to develop and implement strategic focused in optimize the remanent lifetime of the plants. Nevertheless, it is necessary continue working in this field, to achieve a more global and reliable surveillance of the components aging,. In this sense. Tecnatom have among its main objectives, optimize, validate and develop new techniques of degradation detection and diagnosis, improve the remanent lifetime evaluation methods and design surveillance integrated tools, which last achievements have been described in this paper. (Author)

  15. Piezoelectric and Sensitivity Evaluation of Acoustic Emission Sensors for Nuclear Power Plant Valve

    International Nuclear Information System (INIS)

    Lee, Sang-Guk; Lee, Sun-Ki; Park, Sung-Keun; Kim, Myung-Ki

    2008-01-01

    A lot of valves are used in the power plant. The operation safety test and the valve inside leak detection are implemented on the valve which has a great impact on the safe operation of the plant. While input and output pressure measurement using a pressure gauge, temperature change and the humidity measurement, and pressure-resistant test are used for the valve leak detection, there are many problems such as the difficulty of the real time measurement at the minute leak situation, complexity of the pressure gauge correction and the process of the pressure measurement, and the reliability of the measured value. Therefore, it is necessary to develop the valve leak detection system using the acoustic emission (AE) method which is fast and accurate, and allows the real time measurement and evaluation of the minute leak situation. The valve leak detection method using the AE method is a convenient way to detect the sound of the leak outside the valve in case of existing leak inside of the valve, and the research is in progress recently to apply the method to the power plant valve

  16. Nuclear power plant

    International Nuclear Information System (INIS)

    Schabert, H.P.

    1976-01-01

    A nuclear power plant is described which includes a steam generator supplied via an input inlet with feedwater heated by reactor coolant to generate steam, the steam being conducted to a steam engine having a high pressure stage to which the steam is supplied, and which exhausts the steam through a reheater to a low pressure stage. The reheater is a heat exchanger requiring a supply of hot fluid. To avoid the extra load that would be placed on the steam generator by using a portion of its steam output as such heating fluid, a portion of the water in the steam generator is removed and passed through the reheater, this water having received at least adequate heating in the steam generator to make the reheater effective, but not at the time of its removal being in a boiling condition

  17. Nuclear power plant

    International Nuclear Information System (INIS)

    Uruma, Hiroshi

    1998-01-01

    In the first embodiment of the present invention, elements less activated by neutrons are used as reactor core structural materials placed under high neutron irradiation. In the second embodiment of the present invention, materials less activated by neutrons when corrosive materials intrude to a reactor core are used as structural materials constituting portions where corrosion products are generated. In the third embodiment, chemical species comprising elements less activated by neutrons are used as chemical species to be added to reactor water with an aim of controlling water quality. A nuclear power plant causing less radioactivity can be provided by using structural materials comprising a group of specific elements hardly forming radioactivity by activation of neutrons or by controlling isotope ratios. (N.H.)

  18. Wuergassen nuclear power plant

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    The decision of the Federal Court of Administration concerns an application for immediate decommissioning of a nuclear power plant (Wuergassen reactor): The repeal of the permit granted. The decision dismisses the appeal for non-admission lodged by the plaintiffs against the ruling of the Higher Court of Administration (OVG) of North-Rhine Westphalia of December 19th 1988 (File no. 21 AK 8/88). As to the matter in dispute, the Federal Court of Administration confirms the opinion of the Higher Court of Administration. As to the headnotes, reference can be made to that decision. Federal Court of Administration, decision of April 5th 1989 - 7 B 47.89. Lower instance: OVG NW, Az.: 21 AK 8/88. (orig./RST) [de

  19. Evaluation of the control system checkout test at 100% power for Yonggwang Nuclear Power Plant Unit 3

    International Nuclear Information System (INIS)

    Kim, Shin Whan; Lee, Joo Han; Baek, Jong Man; Seo, Jong Tae; Lee, Sang Keun; Kang, In Koo; Ju, Hee Wan; Min, Kyung Soo; Kim, Byung Gon

    1995-01-01

    Control system checkout tests at various powers for Yonggwang Nuclear Power Plant Unit 3(YGN3) were performed to demonstrate the accuracies and proper performances of the control systems of the plant. Tested control systems included the feedwater control system, steam bypass control system, reactor regulation system, control element drive mechanism control system, pressurizer level control system, and pressurizer pressure control system. The measured test data during the control system checkout test at 100% power are evaluated. The test results showed that the control systems of YGN 3 properly control system was simulated by using the LTC code which is the performance analysis code for YGN 3 and 4 design. Comparisons of the predicted results with the measured data confirmed that the feedwater control system controls the steam generator level as designed

  20. A scoping evaluation of severe accidents at Surry and Grand Gulf Nuclear Power Plants resulting from earthquakes during shutdown conditions

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Davis, P.R.

    1991-01-01

    This report explores the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions at two nuclear power plants, Surry Unit I and Grand Gulf Unit 1. The effort is scoping in character, and has been performed primarily to establish if a potential problem exists sufficient to justify a more rigorous and more quantitative evaluation. A summary is presented of the important conclusions that have been reached. The most important conclusion is that the core-damage frequencies for earthquake-initiated accidents during shutdown at both Surry Unit I and Grand Gulf Unit I are found to be low in absolute terms. The reasons for this are that in their ability to respond to earthquakes during shutdowns, the plants both have large seismic capacities, well above their design-basis levels; and also that both sites enjoy among the lowest seismic hazards of any LWR sites in the US

  1. Evaluation of modal properties of cabinet type instrument of nuclear power plant

    International Nuclear Information System (INIS)

    Cho, Y. H.; Park, H. K.; Cho, S. K.

    1999-01-01

    The seismic qualification of safety-related equipment is usually achieved through analysis and testing. Analysis method is preferably adopted for structurally simple equipments which are easy to be mathematically modeled. However, even for relatively complex equipments, analysis method is occasionlly used for computing the input motion or supporting information for the component test followed. Electrical cabinet is a typical example for which analysis method is combinedly used with test to get modal properties of the enclosing cabinet structure. In this paper, with respect to a typical cabinet-type structure(instrumentation cabinet of nuclear power plant) a comparative study has been performed between three different state-of-the-art modeling techniques: lumped mass model, frame model, and FEM modal. From the study results, it has been found that the modal properties of the cabinet-type structure in the elastic behavior range can be reasonably computed through any type of modeling techniques in the practice with slight modification of model properties to get better accuracy. However, it needs additional modeling techniques to get reasonable results up to nonlinear range

  2. Team interaction skills evaluation criteria for nuclear power plant control room operators

    International Nuclear Information System (INIS)

    Montgomery, J.C.; Hauth, J.T.

    1991-01-01

    Team interaction skills are an essential aspect of safe nuclear power plant control room operations. Previous research has shown that, when a group works together, rather than as individuals, more effective operations are possible. However, little research has addressed how such team interaction skills can be measured. In this study rating scales were developed specifically for such a measurement purpose. Dimensions of team skill performance were identified from previous research and experience in the area, incorporating the input of Pacific Northwest Laboratory (PNL) contract operator licensing examiners. Rating scales were developed on the basis of these dimensions, incorporating a modified Behaviorally Anchored Rating Scale (BARS) as well as Behavioral Frequency formats. After a pilot-testing/revision process, rating data were collected using 11 control room crews responding to simulator scenarios at a boiling water and a pressurized water reactor. Statistical analyses of the resulting data revealed moderate inter-rater reliability using the Behavioral Frequency scales, relatively low inter-rater reliability using the BARS, and moderate support for convergent and discriminant validity of the scales. It was concluded that the scales show promise psychometrically and in terms of user acceptability, but that additional scale revision is needed before field implementation. Recommendations for scale revision and directions for future research were presented

  3. Preliminary evaluation of licensing issues associated with U. S. -sited CANDU-PHW nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    van Erp, J B

    1977-12-01

    The principal safety-related characteristics of current CANDU-PHW power plants are described, and a distinction between those characteristics which are intrinsic to the CANDU-PHW system and those that are not is presented. An outline is given of the main features of the Canadian safety and licensing approach. Differences between the U.S. and Canadian approach to safety and licensing are discussed. Some of the main results of the safety analyses, routinely performed for CANDU-PHW reactors, are presented. U.S.-NRC General Design Criteria are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to its conformance to the U.S.-NRC General Design Criteria. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S.

  4. Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Analia Bonelli

    2012-01-01

    Full Text Available A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.

  5. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1984-07-01

    Supplement 27 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for a license to operate Diablo Canyon Nuclear Power Plant, Unit 1 (Docket No. 50-275), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses the revisions to the license conditions and to the Technical Specifications as they relate to Amendment 10 to Diablo Canyon, Unit 1 Facility Operating License, DPR-76

  6. An Evaluation for the Development of 4-channel RSPT and its Application for the OPR1000 Nuclear Power Plants

    International Nuclear Information System (INIS)

    Han, Suk Gyu; Kim, Hyun Min; Song, In Ho; Lee, Yoon Hee; Jeong, Seock Hwan; Kim, Ji Hyeon; Kim, Jae Hack; Kim, Jae Hack; Kim, Young Kook; Kong, Seung Joo

    2008-01-01

    An evaluation project for the development and adaptation of 4-channel reed switch position transmitter (RSPT) has been performed by Korea Power Eng. Company, Inc. (KOPEC) as a contractor of Korea Hydro and Nuclear Power Co. Ltd (KHNP). The 4-channel RSPT is to replace the 2-channel RSPT which is currently installed for all the Optimized Power Reactor 1000 (OPR1000) type nuclear power plants (NPP). The 2-channel RSPT design of OPR1000 could lead an unwanted reactor trip caused by the deviation of a 12-finger control element assembly (CEA) position at each RSPT channel. In addition, the inconsistent channel numbers between 4-channel core protection calculator system (CPCS) and 2-channel RSPT made the system configuration and interface design of the CPCS overly complex. Thus, the 4-channel RSPT development is needed to enhance the OPR1000 plant safety and availability. In this project, while maintaining the existing OPR1000 interfacing system boundary, the 4-channel RSPT manufacturability and the proposed CPCS design with the 4-channel RSPT have been evaluated for their implementation feasibility. A reliability analysis of the proposed CPCS has been also performed. Algorithm changes and the effect of design change regarding interfacing components are also suggested

  7. Evaluation of catch-per-unit-effort indices used in aquatic monitoring programs at nuclear-power-plant sites

    International Nuclear Information System (INIS)

    McKenzie, D.H.; Skalski, J.R.; Simmons, M.A.; Metzger, R.M.

    1982-07-01

    Catch-per-unit-effort (CPUE) indices, or the catch of fish (in numbers or weight) per defined unit of fishing effort, have been used as a fisheries management tool and in monitoring programs at nuclear power plants. An examination of CPUE techniques was conducted with the purpose of developing guidelines for evaluating monitoring programs and interpreting the resulting information. Data bases were selected for analysis from power plant monitoring programs which: (1) had an extensive data base of at least three years; (2) made use of several fish sampling methods; and (3) had monitoring designs which incorporated replicate samples from the fish communities. Two approaches to analyzing and evaluating data bases were taken. At riverine sites, the emphasis centered on applied biological and field sampling; quantitative aspects were evaluated with graphical presentations and coefficients of variation. At the remaining sites, emphasis was placed on statistical aspects and a posteriori sampling designs and hypotheses. Data analyzed from selected programs do not provide evidence to support basic assumptions of CPUE such as proportionality of CPUE to population abundance, sufficiency of CPUE to detect changes of reasonable magnitude, and capability to assess power plant-induced change through CPUE measurement. The findings indicate that CPUE indices cannot be relied on as the sole bases for assessing population changes. Future approaches should be based on a realistic framework that integrates qualitative and quantitative components and recognizes the shortcomings of CPUE as a monitoring tool

  8. Nuclear Security for Floating Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Skiba, James M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scherer, Carolynn P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-13

    Recently there has been a lot of interest in small modular reactors. A specific type of these small modular reactors (SMR,) are marine based power plants called floating nuclear power plants (FNPP). These FNPPs are typically built by countries with extensive knowledge of nuclear energy, such as Russia, France, China and the US. These FNPPs are built in one country and then sent to countries in need of power and/or seawater desalination. Fifteen countries have expressed interest in acquiring such power stations. Some designs for such power stations are briefly summarized. Several different avenues for cooperation in FNPP technology are proposed, including IAEA nuclear security (i.e. safeguards), multilateral or bilateral agreements, and working with Russian design that incorporates nuclear safeguards for IAEA inspections in non-nuclear weapons states

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

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

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

    International Nuclear Information System (INIS)

    1997-06-01

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

  11. Operation reports of nuclear power plants

    International Nuclear Information System (INIS)

    1983-01-01

    The requirements aiming to standardize the program of nuclear power plant operation report, required by Brazilian Energy Commission - CNEN - to evaluate the activities related to the nuclear technical safety and to the radiation protection during the units operational phase, are showed. (E.G.) [pt

  12. Selection procedures for nuclear power plant personnel

    International Nuclear Information System (INIS)

    Kuchler, R.

    1976-01-01

    Selection procedures in reference to experience in staffing two Wisconsin Electric Power Company nuclear project offices and the Point Beach Nuclear Power Plant are discussed. Wisconsin Electric has had a great deal of experience in the application of psychological tests and evaluation procedures, and it was natural that a major consideration in staffing these facilities was the selection of testing procedures

  13. Technological evaluation for the extension of the operation license to the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Arganis J, C. R.; Medina A, A. L.

    2010-01-01

    At the present time one of the tendencies in the nuclear industry is the renovation of operation licenses of the nuclear power plants, with the purpose of prolonging their operation 20 years more than the time settled down in their original license, which is of 30 years for the case of the nuclear power plant of Laguna Verde. This allows the electric power generation for a major period of time and to a relatively low price, giving this way a bigger competitiveness to the power stations of nuclear power. However, to request the license extension of the nuclear power plant requires to get ready the documentation and necessary studies for: to maintain a high level of security, to optimize the operation, maintenance and service life of the structures, systems and components, to maintain an acceptable level of performance, to maximize the recovery of the investment about the service of the nuclear power plant and to preserve the sure conditions for a major operation period at the license time. This paper describes the studies conducted by the Materials Technology Department of the Instituto Nacional de Investigaciones Nucleares (ININ) to substantiate the required documentation for obtaining the extension of operating license of the nuclear power plant. These studies are focused mainly in the reactor pressure vessels of both units, as well as in the deposit of noble metals and the influence of the sludges (crud s) in this deposit. (Author)

  14. Elecnuc. Nuclear power plants worldwide

    International Nuclear Information System (INIS)

    1998-01-01

    This small folder presents a digest of some useful information concerning the nuclear power plants worldwide and the situation of nuclear industry at the end of 1997: power production of nuclear origin, distribution of reactor types, number of installed units, evolution and prediction of reactor orders, connections to the grid and decommissioning, worldwide development of nuclear power, evolution of power production of nuclear origin, the installed power per reactor type, market shares and exports of the main nuclear engineering companies, power plants constructions and orders situation, evolution of reactors performances during the last 10 years, know-how and development of nuclear safety, the remarkable facts of 1997, the future of nuclear power and the energy policy trends. (J.S.)

  15. Conformation of an evaluation process for a license renovation solicitude of a nuclear power plant in Mexico

    International Nuclear Information System (INIS)

    Serrano R, M. L.

    2012-10-01

    So that the construction stages, of operation, closing, dismantlement and the radioactive waste disposal of a nuclear power plant (NPP) are carried out in Mexico, is necessary that the operator has a license, permission or authorization for each stage. In Mexico, these licenses, permissions or authorizations are granted by the Energy Secretariat with base in the verdict of the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). The operation licenses ar the moment effective for the reactors of the Nuclear Power Plant of Laguna Verde (NPP-L V) they will expire respectively in the year 2020 and 2025 for the Unit 1 and Unit 2, for what the CNSNS has begun its preparation before a potential solicitude of the licensee to continue the operation of the NPP-L V. Defining the process to continue and to generate the documents that would help in this phase as normalization, guides, procedures, regulations, controls, etc., is the task that intends to be carried out the regulator body so that the evaluation process is effective and efficient, so much for the same regulator body as for the licensee. This work exposes the advance that the CNSNS has in this aspect and is centered specifically in the conformation of an evaluation process of license renovation solicitude, taking as base what the regulator body of the United States of North America (US NRC) established and following to the IAEA. Also, this work includes statistical of electric power production in Mexico, licensing antecedents for the NPP-L V, a world perspective of the license renovations and the regulation of the US NRC related to the license renovation of a NPP. (Author)

  16. Technical evaluation of RETS-required reports for the Edwin I. Hatch Nuclear Plant, Units 1 and 2

    International Nuclear Information System (INIS)

    Young, T.E.; Magleby, E.H.

    1985-01-01

    A review of the reports required by federal regulations and the plant-specific Radiological Effluent Technical Specifications (RETS) for operations conducted during 1983 was performed. The periodic reports reviewed for the Edwin I. Hatch Nuclear Plant were the Annual Radiological Environmental Operating Report for 1983 and the Semiannual Radioactive Effluent Release Reports for 1983. The principal review guidelines were the plant's specific RETS, NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants'', and NRC Guidance on the Review of the Process Control Programs. The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines. 7 refs

  17. Approval tests in a nuclear power plant

    International Nuclear Information System (INIS)

    Barreto, V.; Rolf, F.

    1984-01-01

    The supplier contract of a Nuclear Power plant establishes the values of Netto electrical Power, specific heat rate and the steam humidity at the outlet of the steam generator. Based on the Nuclear Plant Angra 2 and 3 (1300 MW) the methodology of evaluation and execution of the tests is shown as well as the location of the measuring instruments and the results of the measurements including tolerances. (Author) [pt

  18. Evaluation of the integrity and duration of the Laguna Verde nuclear power plant life- Plant Life Management program (PLIM). TC MEX 04/53 Technical Cooperation Project

    International Nuclear Information System (INIS)

    Arganis J, C.R.; Diaz S, A.; Aguilar T, J.A.

    2006-01-01

    As part of the IAEA TC MEX 04/53 Project 'Evaluation of the integrity and extension of life of the Laguna Verde nuclear power plant Handling Program of plant' whose objective is the one of beginning the actions to apply the methodology of Handling of plant life in the Unit 1 of the Laguna Verde Nucleo electric Central for to obtain the Renovation of License in 2020 the ININ, through the Department of Synthesis and Characterization of materials has carried out more of 20 analysis of susceptibility to the intergranular cracking for corrosion under effort in interns so much of the reactor of the unit 1 like of the unit 2 documenting the current state of components based on the type or types of materials that conform them, to it thermomechanical history, operational and of production, as well as of the particularities associated to its use and operation. For the application of the methodology of life handling of plant 5 structure systems or pilot components were selected, to carry out the programs of handling of the aging and handling of plant life: The encircling of the reactor core (Core Shroud), the reactor pressure vessel (Reactor Pressure Vessel), the primary container (Primary Containment), the recirculation system of feeding water (Reactor Feed Water) and cables. (Author)

  19. Probabilistic safety evaluation: Development of procedures with applications on components used in nuclear power plants

    International Nuclear Information System (INIS)

    Dillstroem, P.

    2000-12-01

    A probabilistic procedure has been developed by SAQ Kontroll AB to calculate two different failure probabilities, P F : Probability of failure, defect size given by NDT/NDE. Probability of failure, defect not detected by NDT/NDE. Based on the procedure, SAQ Kontroll AB has developed a computer program PROPSE (PRObabilistic Program for Safety Evaluation). Within PROPSE, the following features are implemented: Two different algorithms to calculate the probability of failure are included: Simple Monte Carlo Simulation (MCS), with an error estimate on P F . First-Order Reliability Method (FORM), with sensitivity factors using the most probable point of failure in a standard normal space. Using these factors, it is possible to rank the parameters within an analysis. Estimation of partial safety factors, given an input target failure probability and characteristic values for fracture toughness, yield strength, tensile strength and defect depth. Extensive validation has been carried out, using the probabilistic computer program STAR6 from Nuclear Electric and the deterministic program SACC from SAQ Kontroll AB. The validation showed that the results from PROPSE were correct, and that the algorithms used in STAR6 were not intended to work for a general problem, when the standard deviation is either 'small' or 'large'. Distributions, to be used in a probabilistic analysis, are discussed. Examples on data to be used are also given

  20. 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.)

  1. Probabilistic evaluation of risks associated with aviation, road, railway and river traffic adjacent to nuclear power plants in France

    International Nuclear Information System (INIS)

    Procaccia, H.; Gobert, T.

    1978-01-01

    Electricite de France has developed a methodology to calculate the potential impact of risks to a nuclear power plant subsequent to an accident occurring on transportation networks in the vicinity of a nuclear site. Risks entailed in aviation, and in road, rail, and river traffic have been estimated. This methodology has been applied to all sites planned in France

  2. Criteria for preparation and evaluation of radiological emergency response plans and preparedness in support of nuclear power plants. Interim report

    International Nuclear Information System (INIS)

    1980-01-01

    The purpose of this document is to provide a common reference and interim guidance source for: state and local governments and nuclear facility operators in the development of radiological emergency response plans and preparedness in support of nuclear power plants; and Nuclear Regulatory Commission (NRC), Federal Emergency Management Agency (FEMA) and other Federal agency personnel engaged in the review of state, local government, and licensee plans and preparedness

  3. Public regulation of nuclear plants

    International Nuclear Information System (INIS)

    Burtheret, M.; Cormis, de

    1980-01-01

    The construction and operation of nuclear plants are subject to a complex system of governmental administration. The authors list the various governmental authorisations and rules applicable to these plants. In the first part, they describe the national regulations which relate specifically to nuclear plants, and emphasize the provisions which are intended to ensure the safety of the installations and the protection of the public against ionizing radiation. However, while the safety of nuclear plants is a major concern of the authorities, other interests are also protected. This is accomplished by various laws or regulations which apply to nuclear plants as well as other industrial installations. The duties which these texts, and the administrative practice based thereon, impose on Electricite de France are covered in the second part [fr

  4. A dynamic failure evaluation of a simplified digital control system of a nuclear power plant pressurizer

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, J.M.O.; Melo, P.F. Frutuoso e, E-mail: jpinto@nuclear.ufrj.b, E-mail: frutuoso@nuclear.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil); Saldanha, P.L.C., E-mail: plsaldanha@gmail.co, E-mail: pedroabeu@gmail.co [Associacao Brasileira de Ensino Universitario (UNIABEU), Nova Iguacu, RJ (Brazil)

    2010-07-01

    Given the increasing use of digital systems in nuclear power plants, a specific approach to reliability and risk analysis has been required. The digital system reflects many interactions between hardware, software, process variables, and human actions. At the same time, the software, does not have a reliability approach as well-defined as the one existing for the other physical components of the system. Then, its reliability analysis is still under development due to difficulties arising from the complexity, flexibility and interactions present in such systems.The traditional approach of using fault trees is static and does not approach the dynamic interactions in such systems, such as delays in capture and processing information, memory, logic loops, system states, etc. It is necessary to find a reliability methodology that takes into account these issues without violating the existing requirements concerning safety analysis, such as: ability to distinguish between common-cause failures, availability of relevant information to users, like minimal cut sets, and failure probabilities as long as the possibility of incorporating the results into existing probabilistic safety assessments (PSA).One approach is to trace all the possible errors of the digital system through dynamic methodologies. The DFM (Dynamic Flow-graph Methodology) is one of the methodologies that better meets the requirements for modeling dynamic systems. It discretizes the most relevant variables of the analyzed system in states that reflect their behavior, sets the logic that connects them through decision tables and finally performs a system analysis, aiming, for example, the root causes (prime implicants) of a given top event of failure. Three aspects have been addressed, the modeling of the system itself, the incorporation of results to probabilistic safety analyses and identification of software failures.To illustrate the DFM, a simplified digital control system of a typical PWR pressurizer

  5. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    Wood, R.S.

    1991-07-01

    This report indicates percentage ownership of commercial nuclear power plants by utility companies. The report includes all plants operating, under construction, docketed for NRC safety and environmental reviews, or under NRC antitrust review, but does not include those plants announced but not yet under review or those plants formally cancelled. Part 1 of the report lists plants alphabetically with their associated applicants or licensees and percentage ownership. Part 2 lists applicants or licensees alphabetically with their associated plants and percentage ownership. Part 1 also indicates which plants have received operating licenses (OLS)

  6. A computational model for evaluating the effects of cognitive factors on situation assessment of nuclear power plant operators

    International Nuclear Information System (INIS)

    Lee, Hyun Chul

    2010-02-01

    Operators in nuclear power plants have to acquire information from human system interfaces (HSIs) and the environment in order to create, update, and confirm their understanding of a plant state, as failures of situation assessment may cause wrong decisions for process control and finally errors of commission in nuclear power plants. A few computational models that can be used to predict and quantify the situation awareness of operators have been suggested. However, these models do not sufficiently consider human characteristics for nuclear power plant operators. In this thesis, a computational model for situation assessment of nuclear power plant operators using a Bayesian network is proposed. This model incorporates human factors significantly affecting operators' situation assessment, such as attention, working memory decay, and mental model. It is assumed that both the salience level and the information value of information sources are the most significant factors that decide which information sources operators heed. The perception index (PI) is defined as the geometric mean of the salience level and the information value of information sources in this study. The attention allocation rule in the proposed model is defined to direct the operator's attention to the information source with the largest PI value at that moment. There are two approaches for incorporating a mental model of nuclear power plant operators into a Bayesian network: the deterministic rules and the probabilistic rules. The proposed model adopted the probabilistic rules because the deterministic rules are a special case of the probabilistic rules and the probabilistic rules can describe various skill levels of operators. Two working memory decay mechanisms have been suggested: a power law and an exponential law. The proposed model assumes that the situation awareness stored in the working memory decays exponentially because the exponential decay formulas are more popular and require just one

  7. Decision no. 2011-DC-0219 of the French nuclear safety authority from May 5, 2011, ordering the SOCATRI company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the SOCATRI company, operator of the nuclear dismantling and waste processing plants of the Tricastin site (France). (J.S.)

  8. Evaluation of Component Failure Data of the Operating Nuclear Power Plants in Korea Based on NUREG/CR-6928

    International Nuclear Information System (INIS)

    Jeon, Hojun; Na, Janghwan; Shin, Taeyoung

    2014-01-01

    This paper focuses on ensuring the quality of component failure data. When performing data analysis in PSA, we have customized the component failure data based on Bayesian analysis using plant specific experiences and the generic data of Advanced Light Water Reactor Utility Requirements Document (ALWR URD). However, ALWR URD was established by collecting US nuclear power plant (NPP) practices from mid 1980s to early 1990s. We analyzed the component failure data using the raw data of component failures in Pressurized Water Reactor (PWR) plants by 2012. This paper presents the results from analyzing the component failure data based on the new generic data and the latest specific failure data. We also compare the new component failure data to the existing data of PSA models, and evaluate the risk impacts by applying the new data to the PSA models of reference NPPs in this paper. To apply the new generic data source to PSA models, we reviewed and compared NUREG/CR-6928 and the existing generic data source, ALWR URD. In addition, we analyzed the component failure data generated from 16 PWR plants by the end of 2012, and performed the Bayesian update with these raw data based on the new generic data source of NUREG/CR-6928. Also, we reviewed the PSA models of the reference NPP, and identified some important components to CDF. The failure data of the major components decreased in general by applying the new generic data and the latest plant specific data. As a result, the CDF of the reference NPP decreased over 30% compared to the value of the existing CDF

  9. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Salem nuclear power plant, Unit 1

    International Nuclear Information System (INIS)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation is presented for the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Salem nuclear power plant, Unit 1. Design basis criteria used to evaluate the acceptability of the system include operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria

  10. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Yankee Rowe nuclear power plant

    International Nuclear Information System (INIS)

    Latorre, V.R.; Mayn, B.G.

    1979-08-01

    This report documents the technical evaluation of the electrical, instrumentation, and control design aspects for the low temperature overpressure protection system of the Yankee Rowe nuclear power plant. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria

  11. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Maine Yankee nuclear power plant

    International Nuclear Information System (INIS)

    Latorre, V.R.; Mayn, B.G.

    1979-08-01

    This report documents the technical evaluation of the electrical, instrumentation, and control design aspects for the low temperature overpressure protection system of the Maine Yankee nuclear power plant. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria

  12. Analysis and evaluation of the in-pile irradiation test of the fuel element for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Song Wenhui

    1989-10-01

    In order to verify in-pile performances of the fuel element for Qinshan Nuclear Power Plant, the comprehensive in-pile irradiation test of a 3 x 3 - 2 mock fuel element was performed in the Heavy Water Experimental Reactor. Some technical problems and test parameters which have influence on the test results were analyzed in more detail to fully prove that the test is of actual significance. The element was tested under the conditions corresponding to the normal and abnormal conditions of Qinshan NPP. The maximum burnup of the test fuel rods reached 34 GWd/tU and the maximum heat flux of the rods was 1.39 MW/m 2 . The fuel rods didn't fail in the course of test. Finally the test results were evaluated in verifying the performances of the fuel element and its safe operation in Qinshan NPP

  13. CNE (Embalse nuclear power plant): probabilistic safety study. Loss of service water. Probabilistic evaluation and analysis through events sequence

    International Nuclear Information System (INIS)

    Couto, A.J.; Perez, S.S.

    1987-01-01

    This work is part of a study on the service water systems of the Embalse nuclear power plant from a safety point of view. The faults of service water systems of high and low pressure that can lead to situations threatening the plant safety were analyzed in a previous report. The event 'total loss of low pressure service water' causes the largest number of such conditions. Such event is an operational incident that can lead to an accident situation due to faults in the required process systems or by omission of a procedure. The annual frequency of the event 'total loss of low pressure service water' is calculated. The main contribution comes from pump failure. The evaluation of the accident sequences shows that the most direct way to the liberation of fission products is the loss of steam generators as heat sink. The contributions to small and large LOCA and electric supply loss are analyzed. The sequence that leads to tritium release through boiling of moderator is also evaluated. (Author)

  14. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Shiraiwa, Takanori; Watanabe, Atsuo; Miyasawa, Tatsuo

    1984-01-01

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics. (author)

  15. Evaluation of training programs and entry-level qualifications for nuclear-power-plant control-room personnel based on the systems approach to training

    Energy Technology Data Exchange (ETDEWEB)

    Haas, P M; Selby, D L; Hanley, M J; Mercer, R T

    1983-09-01

    This report summarizes results of research sponsored by the US Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research to initiate the use of the Systems Approach to Training in the evaluation of training programs and entry level qualifications for nuclear power plant (NPP) personnel. Variables (performance shaping factors) of potential importance to personnel selection and training are identified, and research to more rigorously define an operationally useful taxonomy of those variables is recommended. A high-level model of the Systems Approach to Training for use in the nuclear industry, which could serve as a model for NRC evaluation of industry programs, is presented. The model is consistent with current publically stated NRC policy, with the approach being followed by the Institute for Nuclear Power Operations, and with current training technology. Checklists to be used by NRC evaluators to assess training programs for NPP control-room personnel are proposed which are based on this model.

  16. Evaluation of training programs and entry-level qualifications for nuclear-power-plant control-room personnel based on the systems approach to training

    International Nuclear Information System (INIS)

    Haas, P.M.; Selby, D.L.; Hanley, M.J.; Mercer, R.T.

    1983-09-01

    This report summarizes results of research sponsored by the US Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research to initiate the use of the Systems Approach to Training in the evaluation of training programs and entry level qualifications for nuclear power plant (NPP) personnel. Variables (performance shaping factors) of potential importance to personnel selection and training are identified, and research to more rigorously define an operationally useful taxonomy of those variables is recommended. A high-level model of the Systems Approach to Training for use in the nuclear industry, which could serve as a model for NRC evaluation of industry programs, is presented. The model is consistent with current publically stated NRC policy, with the approach being followed by the Institute for Nuclear Power Operations, and with current training technology. Checklists to be used by NRC evaluators to assess training programs for NPP control-room personnel are proposed which are based on this model

  17. Computation code TEP 1 for automated evaluation of technical and economic parameters of operation of WWER-440 nuclear power plant units

    International Nuclear Information System (INIS)

    Zadrazil, J.; Cvan, M.; Strimelsky, V.

    1987-01-01

    The TEP 1 program is used for automated evaluation of the technical and economic parameters of nuclear power plant units with WWER-440 reactors. This is an application program developed by the Research Institute for Nuclear Power Plants in Jaslovske Bohunice for the KOMPLEX-URAN 2M information system, delivered by the USSR to the V-2 nuclear power plants in Jaslovske Bohunice and in Dukovany. The TEP 1 program is written in FORTRAN IV and its operation has two parts. First the evaluation of technical and economic parameters of operation for a calculation interval of 10 mins and second, the control of the calculation procedure, follow-up on input data, determination of technical and economic parameters for a lengthy time interval, and data printout and storage. The TEP 1 program was tested at the first unit of the V-2 power plant and no serious faults appeared in the process of the evaluation of technical and economic parameters. A modification of the TEP 1 programme for the Dukovany nuclear power plant is now being tested on the first unit of the plant. (Z.M.)

  18. Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391

    International Nuclear Information System (INIS)

    1992-01-01

    Supplement No. 8 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation of (1) additional information submitted by the applicant since Supplement No. 7 was issued, and (2) matters that the staff had under review when Supplement No. 7 was issued

  19. Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 14

    Energy Technology Data Exchange (ETDEWEB)

    Tam, P.S.

    1994-12-01

    Supplement No. 14 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation with additional information submitted by the applicant since Supplement No. 13 was issued, and matters that the staff had under review when Supplement No. 13 was issued.

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

    International Nuclear Information System (INIS)

    1986-05-01

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

  1. Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, (Docket Nos. 50-390 and 50-391). Supplement No. 2

    International Nuclear Information System (INIS)

    1984-01-01

    This report supplements the Safety Evaluation Report, NUREG-0847 (June 1982) and Supplement No. 1 (September 1982), issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report

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

    International Nuclear Information System (INIS)

    1985-06-01

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

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

    International Nuclear Information System (INIS)

    1986-10-01

    This report, Supplement No. 4 to the Safety Evaluation Report for the application filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicants) for a license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1, 2, and 3

  4. Organizing nuclear power plant operation

    International Nuclear Information System (INIS)

    Adams, H.W.; Rekittke, K.

    1987-01-01

    With the preliminary culmination in the convoy plants of the high standard of engineered safeguards in German nuclear power plants developed over the past twenty years, the interest of operators has now increasingly turned to problems which had not been in the focus of attention before. One of these problems is the organization of nuclear power plant operation. In order to enlarge the basis of knowledge, which is documented also in the rules published by the Kerntechnischer Ausschuss (Nuclear Technology Committee), the German Federal Minister of the Interior has commissioned a study of the organizational structures of nuclear power plants. The findings of that study are covered in the article. Two representative nuclear power plants in the Federal Republic of Germany were selected for the study, one of them a single-unit plant run by an independent operating company in the form of a private company under German law (GmbH), the other a dual-unit plant operated as a dependent unit of a utility. The two enterprises have different structures of organization. (orig.) [de

  5. Evaluation of the epodemiological study on children cancer in the vicinity of nuclear power plants (KiKK study). Scientific justification on the statement of the SSK (Strahlenschutzkommission)

    International Nuclear Information System (INIS)

    Grunst, Marina; Baldauf, Daniela; Reinoehl-Kompa, Sabine

    2008-01-01

    The epidemiological study on children cancer in the vicinity of nuclear power plants (KiKKK) has been en published on December 10, 2007. The study shows that there is a statistical risk for children living close to nuclear power plants to get cancer (leukemia) before the age of 5 years. The Federal Ministry for environment, environmental protection and reactor safety has assigned the SSK (commission for radiation protection) to evaluate this study and to answer the question whether the radiation from nuclear power plants could be responsible fo the observed results. The SSK statement covers the following topics: design and accomplishment of the epidemiological study, actual state of knowledge, independent reevaluation of the KiKKK study, and evaluation of the KiKKK study. A final chapter answers the question of the Ministry

  6. Seismic evaluation of existing nuclear power plants and other facilities V. 1. Proceedings of the technical committee meeting. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    The objectives of this Tcm are: to review the IAEA Safety Report on Seismic Evaluation of Existing Nuclear Power Plants in order to achieve a consensus among Member States on this matter and to discuss the outlines of an IAEA Co-ordinated Research Programme on specific topics related to this subject. Today the nuclear industry relies much more on existing facilities than on the design of new ones. Nevertheless it appears that safety evaluation against external hazards is not a decreasing activity. The reason being that maintaining an acceptable level of nuclear safety requires periodic re-assessments of facilities, either because of modifications of the environment due to human activity or because of new data or new approaches in the assessment of natural hazards. In this regard, seismic re-evaluation has increasingly become a key issue for several existing nuclear facilities, including not only nuclear power plants but also other plants of the fuel cycle, as well as research reactors or laboratories. The IAEA has already supported development of engineering practices in this field by managing a Co-ordinated Research Programme, launched in 1992, on a Benchmark Study for the seismic analysis and testing of WWER 1000 type NPPs. It is now proposed to investigate other aspects of this issue. Many of these facilities were built according to older standards which did not take into account seismic hazard. Consequently, the seismic re-evaluation of existing facilities is a real challenge for earthquake engineers. In most of the cases, it is impossible to re-evaluate according to the up to date standards because entering these standards implies that some design rules are met, what is generally not the case for older facilities. In the best cases some rules exist for non nuclear buildings. In order to achieve a consensus on this matter, the IAEA intends to edit a Safety Report on 'Seismic Evaluation of existing NPPs'. The TCM will offer the opportunity to review the draft of

  7. Partner of nuclear power plants

    International Nuclear Information System (INIS)

    Gribi, M.; Lauer, F.; Pauli, W.; Ruzek, W.

    1992-01-01

    Sulzer, the Swiss technology group, is a supplier of components and systems for nuclear power plants. Important parts of Swiss nuclear power stations, such as containments, reactor pressure vessels, primary pipings, are made in Winterthur. Sulzer Thermtec AG and some divisions of Sulzer Innotec focus their activities on servicing and backfitting nuclear power plants. The European market enjoys priority. New types of valves or systems are developed as economic solutions meeting more stringent criteria imposed by public authorities or arising from operating conditions. (orig.) [de

  8. Technical evaluation of the proposed deletion of a reactor trip on a turbine trip below 50-percent power for the Beaver Valley nuclear power plant, Unit 1

    International Nuclear Information System (INIS)

    Reeves, W.E.

    1979-12-01

    This report documents the technical evaluation of the Duquesne Light Company's proposed license amendment for the deletion of a reactor trip on a turbine trip below 50% power for the Beaver Valley nuclear power plant, Unit 1. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the US Nuclear Regulatory Commission by Lawrence Livermore Laboratory

  9. Assessment of nuclear power plant siting methods

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.D.; Hobbs, B.F.; Pierce, B.L.; Meier, P.M.

    1979-11-01

    Several different methods have been developed for selecting sites for nuclear power plants. This report summarizes the basic assumptions and formal requirements of each method and evaluates conditions under which each is correctly applied to power plant siting problems. It also describes conditions under which different siting methods can produce different results. Included are criteria for evaluating the skill with which site-selection methods have been applied.

  10. Role of satellite and geophysical data for evaluation of site for nuclear power plants - a case study from Rajasthan, India

    International Nuclear Information System (INIS)

    Chaturvedi, A.K.; Ramesh Babu, V.; Srivastava, P.K.; Sunil, T.C.

    2014-01-01

    Remote sensing imageries and seismotectonic maps were studied in the light of recently published regional gravity data jointly by NGRI, OIL, GSI and SOl (2006) for accessing the structural fabric of the area to locate suitable site for nuclear power plant. Entire analysis of the data sets has been carried out in Geographical Information Systems (GIS) environment. The study area, Rawatbhata falls well within the Vindhyan Basin bounded by the towns Bundi in the North, Chittaurgarh in the West, Jhalwar in the Southeast and Kota in the Northeast. Interpretation of the arcuate shaped gravity low in the southeast reveals the probable extension of Hindoli Group of Rocks below the Vidhyans and Traps. The selected location is sitting on the gravity high. A gravity gradient of 5 mGal is observed within the 10km buffer zone which is marked in yellow on all the maps. Earlier workers interpreted the thickness of the Vindhyan sediments as 6-7km, based on the deep seismic and gravity surveys, conducted along the Kunjer-Nagaur transect. Study of the seismotectonic maps published by GSI doesn't indicate any seismic activity over the decades. NE-SW trending neotectonic faults and few thermal springs are noticed far away from the study area around 100 km NW of Great Boundary Fault (GBF). Numbers of NE-SW, NW-SE trending major and minor lineaments/faults are present in the region. However, analysis of the geophysical data sets doesn't reflect any such lineaments/faults within 25 km radius. Integrated study of the area helped in evaluating the existing nuclear power plant site. (author)

  11. Evaluating bio environmental effects of Bushehr Nuclear Power Plant on water and aquatic organism of Persian Gulf

    International Nuclear Information System (INIS)

    Ayatti, F.

    2000-01-01

    The operation of nuclear power plants is always subjected to emission of some radioactive materials in the form of gaseous, liquids and solids in the environment. The heat from condenser coolant discharged to the sea can have some adverse effects on biological systems as thermal pollution. In this project, the radiation and thermal effects on Bushehr Nuclear Power Plants on aquatic animals in Persian Gulf were studied. The mathematical models for atmospheric dispersion of pollutant and pathways of radioactive materials from air to sea water and from sea to animals and human bodies were considered. some environmental samples from Persian Gulf were measured for radioactivity using high-purity Ge/Li detectors and Gamma-spectroscopy. The results indicates that the erection of B usher Nuclear Power Plants and its operation in the normal operation can have no adverse effects on environment, and also its thermal pollution is of no importance due to low area for coolant discharges

  12. Evaluation of the impact of the Nogent nuclear power plant on the Paris area water supply system

    International Nuclear Information System (INIS)

    Leguy, D.; Retkowsky, Y.

    1990-01-01

    The starting up of a nuclear power plant in NOGENT S/SEINE upstream of the main Paris area surface water intake stations led the Seine Normandy Basin Agency and the main water utilities involved to pay attention to the eventual consequences of this nuclear plant on the drinking water supply system of the area. The river Seine itself represents one half of the drinking water resources of the area, more if the intake from the alluvial water table downstream from the capital is taken into account. A working group, run by the Basin Agency and associating the town of Paris, the Ile de France Waterboard, the Compagnie Generale des Eaux, and the Societe Lyonnaise des Eaux was created in 1983 to study the possible impact of the nuclear plant. The main preoccupations which this study highlighted led to the studies and equipments programme this paper presents. (author)

  13. Management of individual and collective dosimetry at Fessenheim nuclear plant. Evaluation after refueling shutdown

    International Nuclear Information System (INIS)

    Lamarre, D.; Waller, A.

    1980-01-01

    The principle of dosimetry management chosen by Fessenheim nuclear power station was originally consisted of two phases: - an automatic acquisition of individual doses realized by stylodosimeter readers; - a deferred data processing by computer. The whole system has not been used during the shutdown for the first refuelling of unit number one in view of encountered difficulties with perfecting of automatic readers prototype, this last phase has been replaced by a manual acquisition of doses. The dosimetry data processing has two main objects: - supervision of individual dosimetry for people who work in the nuclear power station; - knowledge of doses assigned for each working and equipment. Moreover, a first dosimetric result of the shutdown for refuelling of unit number one, enables to notice the workings which doses are the most important and written in percentage of total doses: regulatory controls: about 19%; - steam generators working: 16%; - working decontamination and making health physics screen (lock chamber) 10% [fr

  14. Decision no. 2011-DC-0222 of the French nuclear safety authority from May 5, 2011, ordering the Comurhex company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to Comurhex company, operator of the Tricastin uranium conversion plant (France). (J.S.)

  15. Decision no. 2011-DC-0223 of the French nuclear safety authority from May 5, 2011, ordering the MELOX SA company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to MELOX SA company, operator of the Melox MOX fuel fabrication plant of Marcoule (France). (J.S.)

  16. Decision no. 2011-DC-0218 of the French nuclear safety authority from May 5, 2011, ordering the EURODIF SA company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the EURODIF SA company, operator of the George Besse I uranium enrichment plant of the Tricastin site (France). (J.S.)

  17. Suggestion of Design Evaluation Plan based on Star Life Cycle to introduce the Information Minimalism Concept of KOREA Nuclear Plant

    Science.gov (United States)

    Jang, Gwi-sook; Lee, Seung-min; Park, Gee-yong

    2018-01-01

    The design of Korea Nuclear Power Plant (NPP) main control rooms (MCR) has been changed to be fully digitalized. Five or six display devices are assigned to each operator in NPP MCR to provide the information of safety parameter and plant status, and various control functions by connecting computerized control devices. Under this circumstance, the distributed displays can induce a dispersion of the operators' attention and increase the workload while conducting monitoring and control tasks efficiently. In addition, to support human operators to reduce their workload and increase the performance, the concepts of the ecological interface design (EID) and the operator-centered design were applied to the design HMI display. However these designs are applied to a limited set of screens and did not differ largely from the traditional HMI design in that the layout of the information is somewhere similar to P&IDs. In this paper, we propose a design evaluation plan based on star life cycle to introduce the information minimalism concept for designing an HMI display.

  18. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

    1994-11-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl x , UAl x -Al and U 3 O 8 -Al cermets, U-5% fissium, UMo, UZrH x , UErZrH, UO 2 -stainless steel cermet, and U 3 O 8 -stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified

  19. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R [and others

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  20. Evaluation of some soil amendments plant productivity under saline conditions using nuclear techniques

    International Nuclear Information System (INIS)

    Aly, E.A.K.

    2004-01-01

    this study was carried out in Wadi Ras Sudr (south Saini government). this location was characterized as poor soil with high salinity (wasteland). in the same time it suffers from shortage of water resources. therefore, we aimed to utilize this soil as well as the saline ground water for introducing it into production systems. the reclamation of virgin poor soil need large efforts and much research, especially plant exposure to salinity which is rapidly followed by a decrease in growth rate. the use of natural organic sources as organic fertilizers improve the growth and yields of plants, and safe the environment from pollution . organic fertilizers (Of) such as green manure (G M) or poultry manure (P M) can be used as nutrient sources for good plant growth, where the soil amendments improve the physical, chemical and biological properties of the soil. economically, the yield improvement and nutrient supply will reflect the potential use of such organic materials. also , phosphorus and/or potassium supplementation separately or in combination with O F (G M and/or P M) improved the growth of both barley and wheat plants under such adverse condition of salinity using 15 N isotope dilution technique

  1. Building of nuclear power plant

    International Nuclear Information System (INIS)

    Saito, Takashi.

    1997-01-01

    A first nuclear plant and a second nuclear power plant are disposed in adjacent with each other in a building for a nuclear reactor. A reactor container is disposed in each of the plants, and each reactor container is surrounded by a second containing facility. A repairing chamber capable of communicating with the secondary containing facilities for both of the secondary containing facilities is disposed being in contact with the second containing facility of each plant for repairing control rod driving mechanisms or reactor incorporated-type recycling pumps. Namely, the repairing chamber is in adjacent with the reactor containers of both plants, and situated between both of the plants as a repairing chamber to be used in common for both plants. Air tight inlet/exit doors are formed to the inlets/exits of both plants of the repairing chamber. Space for the repairing chamber can be reduced to about one half compared with a case where the repairing chamber is formed independently on each plant. (I.N.)

  2. Technical evaluation of RETS-required reports for Crystal River Nuclear Generating Plant, Unit 3, for 1983

    International Nuclear Information System (INIS)

    Magleby, E.H.; Young, T.E.

    1985-01-01

    A review was performed on the reports required by Federal regulations and the plant-specific Radiological Effluent Technical Specifications (RETS) for operations conducted at Florida Power Corporation's Crystal River Nuclear Plant, Unit 3, during 1983. The three periodic reports reviewed were (1) the Effluent and Waste Disposal Semiannual Report, January 1-June 30, 1983, (2) the Effluent and Waste Disposal Semiannual Report, July 1-December 31, 1983, and (3) the Annual Environmental Operating Report, Radiological, 1983. The principal review guidelines were the plant's specific RETS and NRC guidance given in NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants.'' The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines

  3. Dispersion of radioactive material in air and water and consideration of population distribution in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The IAEA issues Safety Requirements and Safety Guides pertaining to nuclear power plants and activities in the field of nuclear energy, on the basis of its Safety Fundamentals publication on The Safety of Nuclear Installations. The present Safety Guide, which supplements the Code on the Safety of Nuclear Power Plants: Siting, concerns the effects of a nuclear power plant on the surrounding region and the consideration of population distribution in the siting of a plant. This Safety Guide makes recommendations on how to meet the requirements of the Code on the Safety of Nuclear Power Plants: Siting, on the basis of knowledge of the mechanisms for the dispersion of effluents discharged into the atmosphere and into surface water and groundwater. Relevant site characteristics and safety considerations are discussed. Population distribution, the projected population growth rate, particular geographical features, the capabilities of local transport networks and communications networks, industry and agriculture in the region, and recreational and institutional activities in the region should be considered in assessing the feasibility of developing an emergency response plan. In the selection of a site for a facility using radioactive material, such as a nuclear power plant, account should be taken of any local features that might be affected by the facility and of the feasibility of off-site intervention, including emergency response and protective actions. This is in addition to the evaluation of any features of the site itself that might affect the safety of the facility. This Safety Guide recommends methods for the assessment of regional and local characteristics. This Safety Guide supersedes four earlier IAEA Safety Guides, namely: Atmospheric Dispersion in Nuclear Power Plant Siting (Safety Series No. 50-SG-S3 (1980)). Site Selection and Evaluation for Nuclear Power Plants with Respect to Population Distribution (Safety Series No. 50-SG-S4 (1980)). Hydrological

  4. Evaluation of the Angra-2 nuclear power plant using a RELAP5-PARCS coupled model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patrícia A.L.; Hamers, Adolfo R.; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria A.F.; Verdú, Gumersindo [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Miró, Rafael, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Departamento de Ingenieria Quimica y Nuclear, Universidad Politécnica de Valencia (Spain)

    2017-07-01

    Studies of complex transients in nuclear reactors have been improved by the use of thermal hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes to possibility simulation of transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronic and reactor thermal hydraulics. In this work, steady state results using the verified model of TH RELAP5 code and the NK PARCS code to predict the Angra 2 reactor behavior have been presented. (author)

  5. Nuclear plant simulation using the Nuclear Plant Analyzer

    International Nuclear Information System (INIS)

    Beelman, R.J.; Laats, E.T.; Wagner, R.J.

    1984-01-01

    The Nuclear Plant Analyzer (NPA), a state-of-the-art computerized safety analysis and engineering tool, was employed to simulate nuclear plant response to an abnormal transient during a training exercise at the US Nuclear Regulatory Commission (USNRC) in Washington, DC. Information relative to plant status was taken from a computer animated color graphics display depicting the course of the transient and was transmitted to the NRC Operations Center in a manner identical to that employed during an actual event. Recommendations from the Operations Center were implemented during on-line, interactive execution of the RELAP5 reactor systems code through the NPA allowing a degree of flexibility in training exercises not realized previously. When the debriefing was conducted, the RELAP5 calculations were replayed by way of the color graphics display, adding a new dimension to the debriefing and greatly enhancing the critique of the exercise

  6. Construct ability Improvement for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dae Soo; Lee, Jong Rim; Kim, Jong Ku [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The purpose of this study was to identify methods for improving the construct ability of nuclear power plants. This study reviewed several references of current construction practices of domestic and overseas nuclear plants in order to identify potential methods for improving construct ability. The identified methods for improving construct ability were then evaluated based on the applicability to domestic nuclear plant construction. The selected methods are expected to reduce the construction period, improve the quality of construction, cost, safety, and productivity. Selection of which methods should be implemented will require further evaluation of construction modifications, design changes, contract revisions. Among construction methods studied, platform construction methods can be applied through construction sequence modification without significant design changes, and Over the Top construction method of the NSSS, automatic welding of RCL pipes, CLP modularization, etc., are considered to be applied after design modification and adjustment of material lead time. (author). 49 refs., figs., tabs.

  7. World nuclear power plant capacity

    International Nuclear Information System (INIS)

    1991-01-01

    This report provides the background information for statistics and analysis developed by NUKEM in its monthly Market Report on the Nuclear Fuel Cycle. The assessments in this Special Report are based on the continuous review of individual nuclear power plant projects. This Special Report begins with tables summarizing a variety of nuclear power generating capacity statistics for 1990. It continues with a brief review of the year's major events regarding each country's nuclear power program. The standard NUKEM Market Report tables on nuclear plant capacity are given on pages 24 and 25. Owing to space limitations, the first year shown is 1988. Please refer to previous Special Reports for data covering earlier years. Detailed tables for each country list all existing plants as well as those expected by NUKEM to be in commercial operation by the end of 2005. An Appendix containing a list of abbreviations can be found starting on page 56. Only nuclear power plants intended for civilian use are included in this Special Report. Reactor lifetimes are assumed to be 35 years for all light water reactors and 30 years for all other reactor types, unless other data or definite decommissioning dates have been published by the operators. (orig./UA) [de

  8. Evaluation method of human-system interface for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Tanji, Junichi [Nuclear Power Engineering Corp. (NUPEC), Tokyo (Japan); Kawai, Jun; Masui, Takao; Ezaki, Ikuko [Mitsubishi Research Inst., Tokyo (Japan)

    2000-10-01

    We have developed HSI evaluation method based on a model of human cognitive processes in order to provide the viewpoint of the evaluation on the operability of interface. The model describes systematically the human error categories of GEMS by Reason. Based on the model together with reference to the other published information such as NUREG-0700Rev.1, the evaluation items for HSI have embodied in the electronic handbook HIBISCUS. The applicability of HIBISCUS have been assessed by evaluating experimental results using a simulator equipped with control panel using CRTs and touch operations. From the results, the usefulness of the handbook has been confirmed. (author)

  9. Evaluation method of human-system interface for nuclear power plants

    International Nuclear Information System (INIS)

    Tanji, Junichi; Kawai, Jun; Masui, Takao; Ezaki, Ikuko

    2000-01-01

    We have developed HSI evaluation method based on a model of human cognitive processes in order to provide the viewpoint of the evaluation on the operability of interface. The model describes systematically the human error categories of GEMS by Reason. Based on the model together with reference to the other published information such as NUREG-0700Rev.1, the evaluation items for HSI have embodied in the electronic handbook HIBISCUS. The applicability of HIBISCUS have been assessed by evaluating experimental results using a simulator equipped with control panel using CRTs and touch operations. From the results, the usefulness of the handbook has been confirmed. (author)

  10. Robots for nuclear power plants

    International Nuclear Information System (INIS)

    Moore, T.

    1985-01-01

    In many industrial applications of robots, the objective is to replace human workers with machines that are more productive, efficient, and accurate. But for nuclear applications, the objective is not so much to replace workers as it is to extend their presence - for example, to project their reach into areas of a nuclear plant where the thermal or radiation environment prohibits or limits a human presence. The economic motivation to use robots for nuclear plant inspection and maintenance is centered on their potential for improving plant availability; a by-product is the potential for reducing the occupational radiation exposure of plant personnel. Robotic equipment in nuclear applications may be divided into two broad categories: single-purpose devices with limited ability to perform different operations, and reprogrammable, multi-purpose robots with some degree of computer-based artificial intelligence. Preliminary assessments of the potential for applying robotics in nuclear power plants - mainly at surveillance and inspection tasks - have been carried out. Future developments are considered

  11. ALARA at nuclear power plants

    International Nuclear Information System (INIS)

    Baum, J.W.

    1991-01-01

    Implementation of the ALARA principle at nuclear power plants presents a continuing challenge for health physicists at utility corporate and plant levels, for plant designers, and for regulatory agencies. The relatively large collective doses at some plants are being addressed through a variety of dose reduction techniques. Initiatives by the ICRP, NCRP, NRC, INPO, EPRI, and BNL ALARA Center have all contributed to a heightened interest and emphasis on dose reduction. The NCRP has formed Scientific Committee 46-9 which is developing a report on ALARA at Nuclear Power Plants. It is planned that this report will include material on historical aspects, management, valuation of dose reduction ($/person-Sv), quantitative and qualitative aspects of optimization, design, operational considerations, and training. The status of this work is summarized in this report

  12. Evaluation of the degradation of the service water system in nuclear plants; Evaluacion de la degradacion del sistema de agua de servicio en plantas nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Salaices A, E. [IIE, Av. Reforma 113, Col. Palmira, Cuernavaca, Morelos (Mexico)

    2003-07-01

    The service water system, the circulation water system, the cooling water system and the protection against fires system so much in nuclear plants as in fossils plants they are being degraded by a wide variety of mechanisms. These mechanisms include microbiologically influenced corrosion, cavitation, erosion-corrosion, erosion by solid particles, corrosion in cracks, stings, general corrosion, galvanic corrosion, sedimentation and obstructions and incrustations in the heat exchangers. In the last years were developed predictive models for the more common degradation forms and were installed in a new application of the CHECWORKS{sup TM} code called Cooling Water Application (CWA). This application of the code provides a new technology that so much nuclear facilities as fossil ones can use to modelling specific systems and to carry out corrosion predictions in each one of its components. Presently work the results of the employment of the CHECWORKS CWA code are described to carry out predictions of 12 different corrosion mechanisms that affect to the service water system of a nuclear plant, as well as the recommendations and options that the plant can to consider to reduce indexes of damages. This work can be used for to optimize inspections to the service water system and it gives the bases for similar changes in other nuclear plants. (Author)

  13. Barnwell Nuclear Fuels Plant applicability study. Volume II. BNFP: utilization alternatives, evaluations, and conclusions

    International Nuclear Information System (INIS)

    1978-04-01

    Descriptions and status of the Barnwell separations facility and related fuel cycle facilities are given. Alternative uses other than reprocessing, evaluation of uses for reprocessing alternatives, resource utilization and its relationship to U.S. security objectives, and evaluation of ownership-management options are discussed

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

    International Nuclear Information System (INIS)

    1983-11-01

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

  15. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1991-06-01

    Supplement 34 to the Safety Evaluation Report for the application by Pacific Gas and Electric Company (PG ampersand E) for licenses to operate Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2 (Docket Nos. 50-275 and 50-323, respectively) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement documents the NRC staff review of the Long-Term Seismic Program conducted by PG ampersand E in response to License Condition 2.C.(7) of Facility Operating License DPR-80, the Diablo Canyon Unit 1 operating license. 111 refs., 20 figs., 31 tabs

  16. Analysis of the stochastic somatic fatalities evaluated in the German Nuclear Power Plant Risk Study

    International Nuclear Information System (INIS)

    Ehrhardt, J.

    1981-10-01

    The number of latent cancer fatalities estimated in the German Risk Study (Phase A) is analyzed with respect to the essential models. Special attention is given to the following topics: (1) Allocation of the number of fatalities to the population living during the nuclear accident and to persons born afterwards. (2) Contributions of the relevant exposure pathways and types of cancer to the late fatalities. (3) Estimation of the number of fatalities expected within and beyond a distance of 540 km from the site (alteration of model assumptions). (4) Estimation of the relative portion of late fatalities caused by radiation doses below the dose limits of the German radiation protection regulations. Additionaly, the influence of a non-linear dose response function similar to WASH 1400 on the latent cancer fatalities is estimated. The above-mentioned results provide conclusions concerning necessary improvements of models in Phase B of the Risk Study. (orig.) [de

  17. An application of modulated poisson processes to aging evaluation for the extension of qualified life of nuclear power plant equipment

    International Nuclear Information System (INIS)

    Saldanha, Pedro Luiz da Cruz

    2003-03-01

    The modulated power law process (MPLP) is a three-parameter stochastic point process model that can be used to describe the failures time of reparable systems. While the power law process implies that a system is exactly on the same condition just after a repair as just before a failure, the MPLP allows for the system to be affected both by failure and repair. For this reason this model is adequate for evaluating the recurrent events that incorporate both time trends and effects of past events such as the renewal type behavior. The purpose of this study is to evaluate the MPLP as a model for the rate of occurrence of failures of a repairable system to decide for an extension of qualified life in the context of licence renewal of a nuclear power plant. Standard statistical techniques, such as the maximum likelihood and linear regression models, are applied to estimate parameters of the MPLP. As a conclusion of the study, the MPLP is adequate for modelling rate of occurrence of failures that are time dependent, and can be used where the aging mechanisms are present in the operation of repairable systems. (author)

  18. Study on integrity evaluation of structures associated with nuclear power plants. Annual safety research report, JFY 2011

    International Nuclear Information System (INIS)

    2012-01-01

    This study intends to refine the seismic safety foundational technologies for integrity evaluation of structures on civil, building, equipment, and piping associated with nuclear power plants, in order to find out causes of the failure induced by earthquakes and prepare enhanced safety measures against such external events. Concretely describing, it carries out refinement of analysis methods related to: 3-dimensional model of reactor building, a closer inspection of ceiling crane behavior under an earthquake vibration, and a non-liner relation of stress-strain on reactor building and roof truss structure. It also performs to consider the floor deformational model of reactor building under analyzing earthquake response and to make supplemental descriptions about the model into the analysis guidance, and to develop the seismic isolation structure reviewing guidance. Moreover it carries out refinement of analysis codes which are used for seismic analysis and stress analysis in the implementation of a governmental administration safety review with consideration of knowledge obtained through a technical supporting to revision procedure of analysis evaluation standards and implementation of cross-check analysis on seismic safety reviewing stage. (author)

  19. Medium-size nuclear plants

    International Nuclear Information System (INIS)

    Vogelweith, L.; Lavergne, J.C.; Martinot, G.; Weiss, A.

    1977-01-01

    CEA (TECHNICATOME) has developed a range of pressurized water reactors of the type ''CAS compact'' which are adapted to civil ship propulsion, or to electric power production, combined possibly with heat production, up to outputs equivalent to 125 MWe. Nuclear plants equipped with these reactors are suitable to medium-size electric networks. Among the possible realizations, two types of plants are mentioned as examples: 1) Floating electron-nuclear plants; and 2) Combined electric power and desalting plants. The report describes the design characteristics of the different parts of a 125 MWe unit floating electro-nuclear plant: nuclear steam system CAS 3 G, power generating plant, floating platform for the whole plant. The report gives attention to the different possibilities according to site conditions (the plant can be kept floating, in a natural or artificial basin, it can be put aground, ...) and to safety and environment factors. Such unit can be used in places where there is a growing demand in electric power and fresh water. The report describes how the reactor, the power generating plant and multiflash distillation units of an electric power-desalting plant can be combined: choice of the ratio water output/electric power output, thermal cycle combination, choice of the gain ratio, according to economic considerations, and to desired goal of water output. The report analyses also some technical options, such as: choice of the extraction point of steam used as heat supply of the desalting station (bleeding a condensation turbine, or recovering steam at the exhaust of a backpressure turbine), design making the system safe. Lastly, economic considerations are dealt with: combining the production of fresh water and electric power provides usually a much better energy balance and a lower cost for both products. Examples are given of some types of installations which combine medium-size reactors with fresh water stations yielding from 10000 to 120000 m 3 per day

  20. 78 FR 53774 - Guide for the Evaluation of Alert and Notification Systems for Nuclear Power Plants, FEMA-REP-10...

    Science.gov (United States)

    2013-08-30

    ... available when FEMA REP-10 was first developed, such as smartphones and social media. Removes technical... health and safety in the emergency planning areas surrounding commercial nuclear power plants. The... document. When FEMA issues a final policy, FEMA will publish a notice of availability in the Federal...

  1. Modifications to nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    control and that the modified configuration conforms to the approved basis for granting a nuclear power plant operating licence. The main purpose of the recommendations concerning changes of management is to give general guidance on performing those changes in such a way that the safety of the plant is not compromised. This Safety Guide deals with the intended modification of structures, systems and components, operational limits and conditions, procedures and software, and the management systems and tools for the operation of a nuclear power plant. The recommendations made cover the whole modification process, from conception to completion. The justification for undertaking modifications is outside the scope of this Safety Guide. The modification and/or refurbishment of nuclear power plants for the purpose of extending the design lifetime could necessitate many major design modifications and special re-evaluation of plant safety (see Ref. [2]), and is therefore outside the scope of this publication. Section 2 gives guidance on general methods for modifications that could be implemented at nuclear power plants. Section 3 identifies the roles and responsibilities of various organizations involved in the modification process. Sections 4 and 5 give guidance on the different types of modification and their assessment in respect of safety aspects, and Section 4 provides guidelines on subsequent categorization. Section 6 deals with aspects of temporary modifications. Sections 7 and 8 give guidance on implementation of different types of modifications. Sections 9, 10 and 11 give basic recommendations on quality assurance, training and management of documentation. Comprehensive guidance on these matters can be found in the appropriate Safety Guides

  2. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    Wood, R.S.

    1979-12-01

    The following list indicates percentage ownership of commercial nuclear power plants by utility companies as of December 1, 1979. The list includes all plants licensed to operate, under construction, docketed for NRC safety and envionmental reviews, or under NRC antitrust review. It does not include those plants announced but not yet under review or those plants formally cancelled. In many cases, ownership may be in the process of changing as a result of antitrust license conditions and hearings, altered financial conditions, changed power needs, and other reasons. However, this list reflects only those ownership percentages of which the NRC has been formally notified

  3. Comparison of methods applicable to evaluation of nuclear power plant technical specifications

    International Nuclear Information System (INIS)

    Cho, N.Z.; Bozoki, G.E.; Youngblood, R.W.

    1986-01-01

    The sample problem devised in this study is to calculate core damage frequencies for different allowed outage times (AOTs) from an accident sequence using the three methods: (1) static fault tree approach, (2) time-dependent unavailability analysis, and (3) Markov analysis. The results obtained in this study give useful insights and lend confidence to the conclusion that the static fault tree approach can be used by NRC for the AOT problem when the quantified higher level measures, e.g., core damage frequency and health risks, turn out to be unequivocally below the criteria (e.g., the safety goals). However, when the higher level measures corresponding to the proposed AOTs are in the range of the criteria, the Markov analysis or the time-dependent unavailability analysis would be attractive to owners of the plants who bear the burden of proof in AOT relaxation. 7 refs., 2 figs

  4. Development of regulatory techniques for operational performance evaluation of nuclear power plants

    International Nuclear Information System (INIS)

    Lee, C. J.; Park, J. Y.; Oh, K. M.

    2006-01-01

    In this study, the state-of-the art for estimating LERF is considered for the regulatory risk-informed decisions. The consideration is mainly focused on (1) the relationship between Level 2 PSA and LERF evaluation methodology, (2) the standard requirements in terms of modeling preparation and the acceptance criteria based on the application capability II of ASME PRA standard, and (3) some pending issues for developing and proposing a simplified LERF model. This study is preliminarily presented and will be updated for establishing detailed evaluation scheme of extended MPAS (multi-purpose probabilistic analysis of safety) model and preparing the technical basis

  5. Study of environment evaluation in control room of nuclear power plant

    International Nuclear Information System (INIS)

    Hashimoto, Syusa; Matsushita, Kinzo

    1991-01-01

    We tried to develop an evaluation method of psychological, physiological, and behavioral influence on operators, by using our control room mock-up. We examined the mock-up by changing the environmental condition to test the various items, thus sought a condition so that operators may work more comfortably and eliminate their careless mistakes. The field of Human Factors dealing with an interior environment is being to study. Therefore, a compound, quantitative evaluation on environment have been seldom studied so far. We except that this study will contribute to a coming research about comprehensive interior environment like offices, houses, etc. (author)

  6. Nuclear plant undergrounding

    International Nuclear Information System (INIS)

    Brown, R.C.; Bastidas, C.P.

    1978-01-01

    Under Section 25524.3 of the Public Resources Code, the California Energy Resources Conservation and Development Commission (CERCDC) was directed to study ''the necessity for '' and the effectiveness and economic feasibility of undergrounding and berm containment of nuclear reactors. The author discusses the basis for the study, the Sargent and Lundy (S and L) involvement in the study, and the final conclusions reached by S and L

  7. The influence and evaluation of different virtual reality presentations for the main control room of nuclear power plant

    International Nuclear Information System (INIS)

    Wang Rouwen; Lin Chiuhsiang Joe; Lin Shiaufeng; Yang Chihwei; Cheng Tsungchieh; Yang Lichen

    2011-01-01

    Certainly, a nuclear power plant (NPP) is a complex system and requires high reliability. Engineering technology plays an important role in NPP that requires complex technical equipment and interfaces in order to achieve public security and working safety. Through training, operators can understand the nuclear power system and further establish the fit between human operators and the system, in order to reduce human errors and to ensure the working safety of the control room of NPP. However, the operator trainings for the control room of NPP are difficult and time-consuming. Virtual control room is thus developed using the virtual reality (VR) technology to help the training process. Presently several researches have developed virtual system for NPP for the purpose of training. However, whether the virtual training system for the control room of NPP can give users realistic immersive context as in the real environment is unknown. Whether these virtual systems are helpful in training performance are yet to be confirmed. For this reason, the control room of Lung-Men NPP of Taiwan was constructed with VR technology in this study in order to compare the performances of two VR representation methods (Desktop VR and Project VR). A searching task was planned in which the operators have to find out the objects appointed by the experimenter in the virtual interface of the main control room. The time to complete the task was collected as dependent variables in this experiment. The subjects have to complete the questionnaire that was developed for evaluating the usability of the virtual interface of MRC after finishing the experiment. The result showed that the performance of the virtual interface of NPP presented by the VR projector was better than the desktop. (author)

  8. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-05-01

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

  10. Integrated plant safety assessment: Systematic Evaluation Program, San Onofre Nuclear Generating Station, Unit 1 (Docket No. 50-206): Final report

    International Nuclear Information System (INIS)

    1986-12-01

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues; (2) a basis for deciding on how these differences should be resolved in an integrated plant review; and (3) a documented evaluation of plant safety. This report documents the review of San Onofre Nuclear Generating Station, Unit 1, operated by Southern California Edison Company. The San Onofre plant is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. This report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license. This report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the draft report issued in April 1985

  11. Risk informed inservice inspection evaluation of the ComEd nuclear plants

    International Nuclear Information System (INIS)

    Fleming, K.N.; Hutchinson, J.J.

    2000-01-01

    In November, 1999 Commonwealth Edison Company began the largest scope project yet performed to implement risk informed in-service inspection programs for piping systems. A contractor team led by ERIN Engineering and Research was selected to perform full scope risk informed evaluations on ten reactor units at five sites including 3 BWR and 2 PWR stations. This risk informed evaluation has applied a number of advancements to RISI technology in key areas including risk quantification of inspection program changes, element selection, treatment of multiple damage mechanisms at a given location, and in streamlining an approach to RISI initially developed by EPRI. The scope of the evaluation covered more than 18,000 ASME Class 1 and 2 welds not counting those that will continue to be inspected as part of augmented programs for flow accelerated corrosion, stress corrosion cracking, and service water systems. When the current ASME Section XI based exams are replaced by the risk informed program, while some new inspection locations will be added, substantial cost savings and person-rem exposures are projected from the elimination of most of the current exams. ComEd expects to recover its investment in the RISI evaluation for each of the 10 reactor units in 1 to 2 refueling outages, from the cost savings of reduced inspections. At the same time, the risk of a severe accident due to pipe ruptures is not expected to exhibit significant changes. (author)

  12. Financing evaluation of the Pu spiking project in the Argentine nuclear power plants

    International Nuclear Information System (INIS)

    Bergallo, J.E.; Barcelo, G.N.

    1990-01-01

    The present work pretends to evaluate the benefits that may be obtained from changing the natural uranium cycle that is presently used instead of that of Pu spiking which consists in changing a few elements with mixed oxides fuel with the plutonium reprocessed from all the fuel elements on the periphery of the reactor, maintaining the actual reactor's operating conditions. (Author) [es

  13. Contribution at the evaluation of safety softwares in nuclear power plants control systems

    International Nuclear Information System (INIS)

    Soubies, B.; Le Meur, M.; Henry, J.Y.; Boulc'h, J.

    1993-06-01

    The introduction of programmable systems such the SPIN (Numerical Integrated Protection System) has conducted at particular dispositions for the conception and the use of such systems. The utilization of such systems until 1983 has conducted at modifications in the maintenance procedures. The new methods used for the N4 project in the evaluation of safety softwares are given in this report

  14. Corrosion evaluation and control of cooling systems of nuclear power plants

    International Nuclear Information System (INIS)

    Kim, U. C.; Sung, K. W.; Na, J. W.

    2002-04-01

    We supplemented a database for evaluation of problems in high temperature corrosion degradation at domestic NPPs, and investigated corrosion mitigation methodologies for modification of water chemistry guidelines and operating conditions for life extension as follows: 1) crevice chemistry evaluation by using CAP code, hide-out return tests in a Crevice Test Apparatus, analysis of water chemistry guideline revision history; 2) analysis on main steam generator tube damages, H 2 -dependent crack propagation rate tests of Alloy 600 CRDM material, SCC test of Ni-based alloy in mid-ranged pH by electrochemical polarization and of reduced sulfur influence by corrosion potential measurement, Pb-and CuO-SCC behavior tests, TiO 2 inhibitor penetration test, SCC evaluation, electrochemical observation; 3) loop tests on erosion-corrosion resistance of piping material with ICr-1/2Mo and 21/4 Cr-1Mo, evaluation of current pH-controlling agents based on thermodynamic analysis of amines; 4) SCC susceptibility and H 2 -embrittlement tests of turbine material(3.5NiCrMoV); 5) analyses of Zn-compound for radioactivity reduction and of hydrolysis equilibrium with adsorption test against resins, CERT tests on PWSCC susceptibility of Alloy 600, evaluation of operation mode-dependent radiation level by using CRUSIM code; 6) solubility calculation of radiolysis products of NH 3 , the pH-controlling agent of small-and medium-sized reactors, temperature estimation for obtaining lowest N 2 solubility and of NH 3 content required for optimum pH control, fatigue crack growth test of SG materials(ASTM Grade 2 Ti)

  15. Model-Based Approach to the Evaluation of Task Complexity in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ham, Dong Han

    2007-02-01

    This study developed a model-based method for evaluating task complexity and examined the ways of evaluating the complexity of tasks designed for abnormal situations and daily task situations in NPPs. The main results of this study can be summarised as follows. First, this study developed a conceptual framework for studying complexity factors and a model of complexity factors that classifies complexity factors according to the types of knowledge that human operators use. Second, this study developed a more practical model of task complexity factors and identified twenty-one complexity factors based on the model. The model emphasizes that a task is a system to be designed and its complexity has several dimensions. Third, we developed a method of identifying task complexity factors and evaluating task complexity qualitatively based on the developed model of task complexity factors. This method can be widely used in various task situations. Fourth, this study examined the applicability of TACOM to abnormal situations and daily task situations, such as maintenance and confirmed that it can be reasonably used in those situations. Fifth, we developed application examples to demonstrate the use of the theoretical results of this study. Lastly, this study reinterpreted well-know principles for designing information displays in NPPs in terms of task complexity and suggested a way of evaluating the conceptual design of displays in an analytical way by using the concept of task complexity. All of the results of this study will be used as a basis when evaluating the complexity of tasks designed on procedures or information displays and designing ways of improving human performance in NPPs

  16. Evaluation of fatigue damage in nuclear power plants: evolution and new tools of analysis; Evaluacion del dano a fatiga en centrales nucleares: evolucion y nuevas herramientas de analisis

    Energy Technology Data Exchange (ETDEWEB)

    Cicero, R.; Corchon, F.

    2011-07-01

    This paper presents new fatigue mechanisms requiring analysis, tools developed for evaluation and the latest trends and studies that are currently working in the nuclear field, and allow proper management referring facilities the said degradation mechanism.

  17. Inventory and source term evaluation of Russian nuclear power plants for marine applications

    International Nuclear Information System (INIS)

    Reistad, O.; Oelgaard, P.L.

    2006-04-01

    This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

  18. Inventory and source term evaluation of Russian nuclear power plants for marine applications

    Energy Technology Data Exchange (ETDEWEB)

    Reistad, O. [Norwegian Radiation Protection Authority (Norway); Oelgaard, P.L. [Risoe National Lab. (Denmark)

    2006-04-15

    This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

  19. Probabilistic evaluation of fire protection features found in nuclear power plants

    International Nuclear Information System (INIS)

    Azarm, M.A.; Boccio, J.L.; Ruger, C.

    1985-01-01

    This paper describes a method which can be used to evaluate, on a relative basis, the NRC Fire Protection (FP) guidelines as found in Section 9.5.1 (Fire Protection) of the Standard Review Plan (SRP). The approach, a hybrid of existing physical models for fire propagation determinations and probabilistic models for fire-mitigation system reliability, can potentially be used as an adjunct to the present fire safety review process

  20. Evaluation of the radioactive waste characterisation at the Olkiluoto nuclear power plant

    International Nuclear Information System (INIS)

    Kekki, T.; Titta, A.

    2000-03-01

    The aim of this study is to evaluate the physical, chemical and radiological characterisation, handling and documentation of the radioactive waste packages to be disposed of in the VLJ-repository at the Olkiluoto NPP. A comparison with the current practices in Europe, based on information from Sweden, Spain and Czech Republic, is made. The report presents recommendations for STUK to harmonise the LILW waste management practises in Finland with those in Europe. (orig.)

  1. Systems (CSS) for nuclear power plants (1993-1995): Evaluation of CSS using simulated human model

    International Nuclear Information System (INIS)

    Yushi Fujita; Kazushige Gotoh

    1996-01-01

    A simulated human model called Cognitive and Action Modelling of Erring Operator (CAMEO) has been developed. CAMEO is designed to be utilized as a platform for evaluating human-machine interfaces including CSS. Based on experimental studies and computer simulation summarized in this paper, it may be possible to conclude that CAMEO is reasonably valid for predicting some typical cognitive difficulties that human operators might experience in supervisory control environments. (author). 6 refs

  2. Evaluated Nuclear Data

    Energy Technology Data Exchange (ETDEWEB)

    Oblozinsky, P.; Oblozinsky,P.; Herman,M.; Mughabghab,S.F.

    2010-10-01

    This chapter describes the current status of evaluated nuclear data for nuclear technology applications. We start with evaluation procedures for neutron-induced reactions focusing on incident energies from the thermal energy up to 20 MeV, though higher energies are also mentioned. This is followed by examining the status of evaluated neutron data for actinides that play dominant role in most of the applications, followed by coolants/moderators, structural materials and fission products. We then discuss neutron covariance data that characterize uncertainties and correlations. We explain how modern nuclear evaluated data libraries are validated against an extensive set of integral benchmark experiments. Afterwards, we briefly examine other data of importance for nuclear technology, including fission yields, thermal neutron scattering and decay data. A description of three major evaluated nuclear data libraries is provided, including the latest version of the US library ENDF/B-VII.0, European JEFF-3.1 and Japanese JENDL-3.3. A brief introduction is made to current web retrieval systems that allow easy access to a vast amount of up-to-date evaluated nuclear data for nuclear technology applications.

  3. Countermeasures to earthquakes in nuclear plants

    International Nuclear Information System (INIS)

    Sato, Kazuhide

    1979-01-01

    The contribution of atomic energy to mankind is unmeasured, but the danger of radioactivity is a special thing. Therefore in the design of nuclear power plants, the safety has been regarded as important, and in Japan where earthquakes occur frequently, the countermeasures to earthquakes have been incorporated in the examination of safety naturally. The radioactive substances handled in nuclear power stations and spent fuel reprocessing plants are briefly explained. The occurrence of earthquakes cannot be predicted effectively, and the disaster due to earthquakes is apt to be remarkably large. In nuclear plants, the prevention of damage in the facilities and the maintenance of the functions are required at the time of earthquakes. Regarding the location of nuclear plants, the history of earthquakes, the possible magnitude of earthquakes, the properties of ground and the position of nuclear plants should be examined. After the place of installation has been decided, the earthquake used for design is selected, evaluating live faults and determining the standard earthquakes. As the fundamentals of aseismatic design, the classification according to importance, the earthquakes for design corresponding to the classes of importance, the combination of loads and allowable stress are explained. (Kako, I.)

  4. Risk-based evaluation of allowed outage time and surveillance test interval extensions for nuclear power plants

    International Nuclear Information System (INIS)

    Gibelli, Sonia Maria Orlando

    2008-03-01

    The main goal of this work is, through the use of Probabilistic Safety Analysis (PSA), to evaluate Technical Specification (TS) Allowed Outage Times (AOT) and Surveillance Test Intervals (STI) extensions for Angra 1 nuclear power plant. PSA has been incorporated as an additional tool, required as part of NPP licensing process. The risk measure used in this work is the Core Damage Frequency (CDF), obtained from the Angra 1 PSA Level 1. AOT and STI extensions are calculated for the Safety Injection System (SIS), Service water System (SAS) and Auxiliary Feedwater System (AFS) through the use of SAPHIRE code. In order to compensate for the risk increase caused by the extensions, compensatory measures as test of redundant train prior to entering maintenance and staggered test strategy are proposed. Results have shown that the proposed AOT extensions are acceptable for the SIS and SAS with the implementation of compensatory measures. The proposed AOT extension is not acceptable for the AFS. The STI extensions are acceptable for all three systems. (author)

  5. Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category I systems for Palisades nuclear power plant

    International Nuclear Information System (INIS)

    Collins, E.K.

    1979-10-01

    The technical evaluation is presented of Consumers Power Company's Palisades nuclear power plant to determine whether the failure of any non-Category I (seismic) equipment could result in a condition, such as flooding, that might potentially adversely affect the performance of safety-related equipment required for the safe shutdown of the facility or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection as well as measures taken by Consumers Power Company to minimize the danger of flooding and to protect safety-related equipment

  6. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach Nuclear Power Plant, Units 1 and 2

    International Nuclear Information System (INIS)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation is presented for the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach nuclear power plant, Units 1 and 2. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Support Program being conducted for the U.S. Nuclear Regulatory Commission by Lawrence Livermore Laboratory

  7. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach Nuclear Power Plant, Units 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation is presented for the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Point Beach nuclear power plant, Units 1 and 2. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Support Program being conducted for the U.S. Nuclear Regulatory Commission by Lawrence Livermore Laboratory.

  8. Technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Haddam Neck Nuclear Power Plant

    International Nuclear Information System (INIS)

    Laudenbach, D.H.

    1979-03-01

    The technical evaluation of the electrical, instrumentation, and control design aspects of the low temperature overpressure protection system for the Haddam Neck Nuclear Power Plant is presented. Design basis criteria used to evaluate the acceptability of the system included operator action, system testability, single failure criterion, and seismic Category I and IEEE Std-279-1971 criteria. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Support Program being conducted for the U.S. Nuclear Regulatory Commission by Lawrence Livermore Laboratory

  9. Structural Aging Program to evaluate continued performance of safety-related concrete structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1994-01-01

    This report discusses the Structural Aging (SAG) Program which is being conducted at the Oak Ridge National Laboratory (ORNL) for the United States Nuclear Regulatory commission (USNRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into three technical tasks: Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for continued Service Determinations. Objectives and a summary of recent accomplishments under each of these tasks are presented

  10. Structural aging program to evaluate continued performance of safety-related concrete structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1994-01-01

    The Structural Aging (SAG) Program is being conducted at the Oak Ridge National Laboratory (ORNL) for the United States Nuclear Regulatory Commission (USNRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into three technical tasks: Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for Continued Service Determinations. Objectives and a summary of recent accomplishments under each of these tasks are presented. (Author)

  11. Elecnuc. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    2003-01-01

    This 2003 version of Elecnuc contents information, data and charts on the nuclear power plants in the world and general information on the national perspectives concerning the electric power industry. The following topics are presented: 2002 highlights; characteristics of main reactor types and on order; map of the French nuclear power plants; the worldwide status of nuclear power plants on 2002/12/3; units distributed by countries; nuclear power plants connected to the Grid by reactor type groups; nuclear power plants under construction; capacity of the nuclear power plants on the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear plants by country at the end 2002; performance indicator of french PWR units; trends of the generation indicator worldwide from 1960 to 2002; 2002 cumulative Load Factor by owners; nuclear power plants connected to the grid by countries; status of license renewal applications in Usa; nuclear power plants under construction; Shutdown nuclear power plants; exported nuclear power plants by type; exported nuclear power plants by countries; nuclear power plants under construction or order; steam generator replacements; recycling of Plutonium in LWR; projects of MOX fuel use in reactors; electricity needs of Germany, Belgium, Spain, Finland, United Kingdom; electricity indicators of the five countries. (A.L.B.)

  12. Decision no. 2011-DC-0216 of the French nuclear safety authority from May 5, 2011, ordering the Laue Langevin Institute to proceed to a complementary safety evaluation of its basic nuclear facility (high flux reactor - INB no. 67) in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the Laue Langevin Institute, operator of the high flux research reactor (RHF) of Grenoble (France). (J.S.)

  13. Decision no. 2011-DC-0215 of the French nuclear safety authority from May 5, 2011, ordering ITER Organization to proceed to a complementary safety evaluation of its basic nuclear facility in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the ITER Organization, operator of the ITER tokamak facility of Cadarache (France). (J.S.)

  14. Decision no. 2011-DC-0224 of the French nuclear safety authority from May 5, 2011, ordering the French atomic energy and alternative energies commission (CEA) to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the French atomic energy commission (CEA). (J.S.)

  15. Decision no. 2011-DC-0214 of the French nuclear safety authority from May 5, 2011, ordering CIS bio international company to proceed to a complementary safety evaluation of its basic nuclear facility in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to CIS bio international company, operator of the radiopharmaceuticals fabrication facility (INB 29) of Saclay (France). (J.S.)

  16. Nuclear plant aging research program

    International Nuclear Information System (INIS)

    Eissenberg, D.M.

    1987-01-01

    The U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, has established the Nuclear Plant Aging Research (NPAR) program in its Division of Engineering Technology. Principal contractors for this program include Oak Ridge National Laboratory, Brookhaven National Laboratory, Idaho National Engineering Laboratory, and Pacific Northwest Laboratory. The program goals are: to identify and characterize time-dependent degradation (aging) of nuclear plant safety-related electrical and mechanical components which could lead to loss of safety function; to identify and recommend methods for detecting and trending aging effects prior to loss of safety function so that timely maintenance can be implemented; and to recommend maintenance practices for mitigating the effects of aging. Research activities include prioritization of system and component aging in nuclear plants, characterization of aging degradation of specific components including identification of functional indicators useful for trending degradation, and testing of practical methods and devices for measuring the functional indicators. Aging assessments have been completed on electric motors, snubbers, motor-operated valves, and check valves. Testing of trending methods and devices for motor-operated valves and check valves is in progress

  17. Underwater nuclear power plant structure

    International Nuclear Information System (INIS)

    Severs, S.; Toll, H.V.

    1982-01-01

    A structure for an underwater nuclear power generating plant comprising a triangular platform formed of tubular leg and truss members upon which are attached one or more large spherical pressure vessels and one or more small cylindrical auxiliary pressure vessels. (author)

  18. evaluation of selected composted organic sources on potato plant grown in sandy soil using nuclear technique

    International Nuclear Information System (INIS)

    Moursy, A.A.A.

    2008-01-01

    the main point of this study is the evaluation of organic compost as a source of nutrient demand by potatoes cultivated in light texture soil under drip irrigation system. the composted materials either applied alone or in combination with mineral fertilizer have an effective role on potato yields and nutrients management under field scale. so, many objectives were achieved. the valuable results obtained in the present study could be summarized as follows: part one: composting experiment contains ph changes of composted materials, EC changes with time, nitrogen content in composted materials, change of c/n ratio with time, organic matter content of the composted materials, phosphorus content in composted materials,. part two: potato field experiment contains .dry matter yield, tuber dry weight, tuber yield, nutrients uptake by potato varieties,. part three contains . application of 15 N isotope dilution technique, nitrogen derived from fertilizer (Ndff), nitrogen derived from organic compost (% Ndf comp),nitrogen derived from soil (% Ndfs), fertilizer use efficiency (% FUE), 15 N recovered by potatoes.

  19. Heuristic guidelines and experimental evaluation of effective augmented-reality based instructions for maintenance in nuclear power plants

    International Nuclear Information System (INIS)

    Yim, Ho Bin; Seong, Poong Hyun

    2010-01-01

    Research highlights: → Augmented reality (AR) instructions were built for NPPs maintenance personnel. → 4-5 pieces of information at a time were optimum for AR instructions in this study. → A large variance in mode no. 5 implies these were also found to be critical amount. → Heuristic guidelines were suggested to make AR instructions more effective. - Abstract: As industrial plants and factories age, their maintenance requirements increase. Because maintenance mistakes directly increase the operating costs of a power plant, maintenance quality is significant concern to plant management. By law, all personnel working with nuclear technology must be re-trained every three years in Korea; however, as the statistical data show, the number of shutdown accidents at nuclear power plants (NPPs) due to maintenance failure is still high and needs to be reduced. Industries have started to adopt various technologies to increase the speed and accuracy of maintenance. Among those technologies, augmented reality (AR) is the latest multimedia presentation technology to be applied to plant maintenance, and it offers superior intuitiveness and user interactivity over other conventional multimedia. This empirical study aims to measure the optimum amounts of information to be delivered at a time and to identify what types of information enhance the learning ability of novices and to suggest heuristic guidelines by which to make effective AR training instructions. In the first experiment, the optimum amount of information in an AR learning environment for novices was found to be 4-5 pieces of information in a chunk by comparing results between a pre-test and an after-test. This result implies that intentionally made chunks help novices learn more effectively. In the second experiment, the AR training instruction based on the suggested heuristic guidelines was slightly more effective than other AR training instructions. Maintenance in nuclear power plants can be more reliable

  20. Development of Nuclear Power Plant Safety Evaluation Method for the Automation Algorithm Application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Geun; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    It is commonly believed that replacing human operators to the automated system would guarantee greater efficiency, lower workloads, and fewer human error. Conventional machine learning techniques are considered as not capable to handle complex situations in NPP. Due to these kinds of issues, automation is not actively adopted although human error probability drastically increases during abnormal situations in NPP due to overload of information, high workload, and short time available for diagnosis. Recently, new machine learning techniques, which are known as ‘deep learning’ techniques have been actively applied to many fields, and the deep learning technique-based artificial intelligences (AIs) are showing better performance than conventional AIs. In 2015, deep Q-network (DQN) which is one of the deep learning techniques was developed and applied to train AI that automatically plays various Atari 2800 games, and this AI surpassed the human-level playing in many kind of games. Also in 2016, ‘Alpha-Go’, which was developed by ‘Google Deepmind’ based on deep learning technique to play the game of Go (i.e. Baduk), was defeated Se-dol Lee who is the World Go champion with score of 4:1. By the effort for reducing human error in NPPs, the ultimate goal of this study is the development of automation algorithm which can cover various situations in NPPs. As the first part, quantitative and real-time NPP safety evaluation method is being developed in order to provide the training criteria for automation algorithm. For that, EWS concept of medical field was adopted, and the applicability is investigated in this paper. Practically, the application of full automation (i.e. fully replaces human operators) may requires much more time for the validation and investigation of side-effects after the development of automation algorithm, and so the adoption in the form of full automation will take long time.

  1. Development of Nuclear Power Plant Safety Evaluation Method for the Automation Algorithm Application

    International Nuclear Information System (INIS)

    Kim, Seung Geun; Seong, Poong Hyun

    2016-01-01

    It is commonly believed that replacing human operators to the automated system would guarantee greater efficiency, lower workloads, and fewer human error. Conventional machine learning techniques are considered as not capable to handle complex situations in NPP. Due to these kinds of issues, automation is not actively adopted although human error probability drastically increases during abnormal situations in NPP due to overload of information, high workload, and short time available for diagnosis. Recently, new machine learning techniques, which are known as ‘deep learning’ techniques have been actively applied to many fields, and the deep learning technique-based artificial intelligences (AIs) are showing better performance than conventional AIs. In 2015, deep Q-network (DQN) which is one of the deep learning techniques was developed and applied to train AI that automatically plays various Atari 2800 games, and this AI surpassed the human-level playing in many kind of games. Also in 2016, ‘Alpha-Go’, which was developed by ‘Google Deepmind’ based on deep learning technique to play the game of Go (i.e. Baduk), was defeated Se-dol Lee who is the World Go champion with score of 4:1. By the effort for reducing human error in NPPs, the ultimate goal of this study is the development of automation algorithm which can cover various situations in NPPs. As the first part, quantitative and real-time NPP safety evaluation method is being developed in order to provide the training criteria for automation algorithm. For that, EWS concept of medical field was adopted, and the applicability is investigated in this paper. Practically, the application of full automation (i.e. fully replaces human operators) may requires much more time for the validation and investigation of side-effects after the development of automation algorithm, and so the adoption in the form of full automation will take long time

  2. Nuclear power plant

    International Nuclear Information System (INIS)

    Amamiya, Shu; Inagaki, Shuichi; Suzuki, Yasuhiro.

    1992-01-01

    The plant of the present invention can suppress the amount of clad in feedwater when drains of a moisture content separation heater or a high pressure feedwater heater are recovered. That is, the moisture content separation heater has ferrite or austenite type stainless steel heat transfer pipes. A chromium-enriched layer is formed on the surface of the heat transfer pipe by chromizing treatment or flame spraying. Then, a stainless steel heat transfer pipe having chromium-enriched layer is incorporated to at least one of the moisture content separation heater or the high pressure feedwater heater. During plant operation, the temperature of heated steams is as high as 235 to 282degC. Accordingly, this is a severe corrosion region for ferrite or austenite stainless steel. However, the chromium-enriched layer of excellent corrosion resistance is formed on the surface of the heat transfer pipe. Accordingly, metal ingredients are less leached. As a result, even if the drains are recovered to feedwater, increase of concentration of the clads in the feedwater can be prevented. (I.S.)

  3. Preliminary study of the emergency planning zone evaluation for the nuclear power plant in Taiwan by using MACCS2 code

    International Nuclear Information System (INIS)

    Lo, Chung-Kung; Chen, Ing-Jane; Huang, Yu-Hao; Chou, Yuan-Ching

    2000-01-01

    According to government regulation, the Emergency Planning Zone (EPZ) of Nuclear Power Plant (NPP) must be designated before operation. The related emergency response planning (ERP) in EPZ must be planned to guarantee all necessary resources are available under the postulated accidents of NPP. Thus the pre-planned necessary actions will be helpful to protect population from the damage during the possible accident. The purpose of this preliminary study is to perform the evaluation of EPZ for a NPP by using MACCS2 (MELCOR Accident Consequence Code System 2) code which was developed by SNL for estimating the radiological doses, health effects, and economic consequences that could result from postulated accidental releases of radioactive materials to the atmosphere. The specific meteorological data used in this study was collected during July 1997 to June 1998 at the planned site and the population distribution was reinvestigated in April 1999. The source term data including inventory, sensible heat content, timing, duration, the fraction of inventory released with each segment etc. are based on a preliminary design result of the NPP. The effective dose equivalent and thyroid dose together with the related individual risk, social risk for each category of accidents are evaluated by MACCS2 code. An interface program has been developed to summarize the dose and risk results from all the accident categories. By comparing the results of individual risk, societal risk, whole body dose, and thyroid dose vs. distance to the Protection Action Guide and the NUREG-0396 report related criteria, a reasonably conservative EPZ is proposed. (author)

  4. Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Suppl. 22

    International Nuclear Information System (INIS)

    1984-03-01

    Supplement 22 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Unit 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the US Nuclear Regulatory Commission. This supplement provides the criteria that were used by the staff to determine which of the allegations that have been evaluated and must be resolved prior to Unit 1 achieving criticality and operating at power level up to 5 percent of rated power (i.e., low power operation). The supplement also reports on the status of the staff's investigation, inspection and evaluation of 219 allegations or concerns that have been identified to the NRC as of March 9, 1984, excluding those recently received under 10 CFR 2.206 petitions

  5. The Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    Dr. David A. Petti

    2009-01-01

    The Next Generation Nuclear Plant (NGNP) will be a demonstration of the technical, licensing, operational, and commercial viability of High Temperature Gas-Cooled Reactor (HTGR) technology for the production of process heat, electricity, and hydrogen. This nuclear- based technology can provide high-temperature process heat (up to 950°C) that can be used as a substitute for the burning of fossil fuels for a wide range of commercial applications (see Figure 1). The substitution of the HTGR for burning fossil fuels conserves these hydrocarbon resources for other uses, reduces uncertainty in the cost and supply of natural gas and oil, and eliminates the emissions of greenhouse gases attendant with the burning of these fuels. The HTGR is a passively safe nuclear reactor concept with an easily understood safety basis that permits substantially reduced emergency planning requirements and improved siting flexibility compared to other nuclear technologies.

  6. BWR type nuclear plant

    International Nuclear Information System (INIS)

    Fujita, Yoshio; Okano, Kimifumi; Sasaki, Hiroshi; Okura, Minoru

    1998-01-01

    The present invention provides a BWR plant capable of reducing the size of the reactor building while maintaining reliability of a pool water cooling and cleaning facility even when two fuel storage pools are disposed in the reactor building. Namely, in the reactor building, two fuel storage pools, a temporality storing pool for temporary storing incore structures and a suppression pool are disposed. A primary cleaning facility for cooling and cleaning pool water for each of fuel storage pools comprises a serge tank, a pump, a heat exchanger and a filtration desalting device. A secondary cleaning facility for cleaning pool water in the suppression pool comprises a pump and a filtration desalting device. The first cleaning facility can be switched to be used for the secondary cleaning facility. Specifically, upstream and downstream of the pump of the primary cleaning facility and those of the pump of the secondary cleaning facility are connected by communication pipelines. (I.S.)

  7. The long view for nuclear plant maintenance

    International Nuclear Information System (INIS)

    Moore, T.

    1991-01-01

    This article discusses a strategic, anticipatory approach to maintenance as the key to the long-term viability of today's nuclear power plants. As many as 20 plants around the country now have life-cycle management (LCM) programs in place - integrated, forward-looking preventive maintenance and monitoring programs that preserve the plant's material condition and extend economical operation. Besides reducing the need for 'pounds of cure' decades in the future, LCM programs can produce significant near-term gains in plant performance and are valuable in addressing license renewal concerns. EPRI is supporting the industry's LCM activities with research on age-related degradation mechanisms, development of economic evaluation tools for applying LCM approaches, and assistance with life-cycle evaluations of major plant systems, structures, and components at selected utilities

  8. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Gary T [ORNL; Belles, Randy [ORNL; Cetiner, Sacit M [ORNL; Howard, Rob L [ORNL; Liu, Cheng [ORNL; Mueller, Don [ORNL; Omitaomu, Olufemi A [ORNL; Peterson, Steven K [ORNL; Scaglione, John M [ORNL

    2012-06-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  9. Investment issues in nuclear plant license renewal

    International Nuclear Information System (INIS)

    Eynon, R.T.

    1999-01-01

    A method that determines the operating lives for existing nuclear power plants is discussed. These assumptions are the basis for projections of electricity supply through 2020 reported in the Energy Information Administration's (EIA's) Annual Energy Outlook 1999. To determine if plants will seek license renewal, one must first determine if they will be operating to the end of their current licenses. This determination is based on an economic test that assumes an investment of $150/kW will be required after 30 yr of operation for plants with older designs. This expenditure is intended to be equivalent to the cost that would be associated with any of several needs such as a one0time investment to replace aging equipment (steam generators), a series of investments to fix age-related degradation, increases in operating costs, or costs associated with decreased performance. This investment is compared with the cost of building and operating the lowest-cost new plant over the same 10-yr period. If a plant fails this test, it is assumed to be retired after 30 yr of service. All other plants are then considered candidates for license renewal. The method used to determine if it is economic to apply for license renewal and operate plants for an additional 20 yr is to assume that plants face an investment of $250 million after 40 yr of operation to refurbish aging components. This investment is compared with the lowest-cost new plant alternative evaluated over the same 20 yr that the nuclear plant would operate. If the nuclear plant is the lowest cost option, it is projected to continue to operate. EIA projects that it would be economic to extend the operating licenses for 3.7 GW of capacity (6 units)

  10. Nuclear power plants: structure and function

    International Nuclear Information System (INIS)

    Hendrie, J.M.

    1983-01-01

    Topics discussed include: steam electric plants; BWR type reactors; PWR type reactors; thermal efficiency of light water reactors; other types of nuclear power plants; the fission process and nuclear fuel; fission products and reactor afterheat; and reactor safety

  11. Relative costs to nuclear plants: international experience

    International Nuclear Information System (INIS)

    Souza, Jair Albo Marques de

    1992-03-01

    This work approaches the relative costs to nuclear plants in the Brazil. It also presents the calculation methods and its hypothesis to determinate the costs, and the nacional experience in costs of investment, operating and maintenance of the nuclear plants

  12. Passive Nuclear Plants Program (UPDATE)

    International Nuclear Information System (INIS)

    Chimeno, M. A.

    1998-01-01

    The light water passive plants program (PCNP), today Advanced Nuclear Power Plants Program (PCNA), was constituted in order to reach the goals of the Spanish Electrical Sector in the field of advanced nuclear power plants, optimize the efforts of all Spanish initiatives, and increase joint presence in international projects. The last update of this program, featured in revision 5th of the Program Report, reflects the consolidation of the Spanish sector's presence in International programs of the advanced power plants on the basis of the practically concluded American ALWR program. Since the beginning of the program , the PCNP relies on financing from the Electrical sector, Ocide, SEPI-Endesa, Westinghouse, General Electric, as well as from the industrial cooperators, Initec, UTE (Initec- Empresarios Agrupados), Ciemat, Enusa, Ensa and Tecnatom. The program is made up of the following projects, already concluded: - EPRI's Advanced Light Water Plants Certification Project - Westinghouse's AP600 Project - General Electric's SBWR Project (presently paralyzed) and ABWR project Currently, the following project are under development, at different degrees of advance: - EPP project (European Passive Plant) - EBWR project (European Advanced Boiling Water Reactor)

  13. Seismic safety of nuclear power plants

    International Nuclear Information System (INIS)

    Guerpinar, A.; Godoy, A.

    2001-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  14. Turbines for nuclear power plants. 2.ed.

    International Nuclear Information System (INIS)

    Troyanovskij, B.M.

    1978-01-01

    In the second edition of the book considered are practically all the main problems of calculation and operation of turbines and turbine installations of nuclear power plants. As compared to the first edition, essentially addes is the reproduction of the problem on combined generation of heat and electric energy. Also represented is detailed material on methods of preliminary evaluation of turbine effectiveness. Considered are peculiarities of turbine operation on wet steam and the basis of their thermal calculation. Much attention is payed to the problem of wet stream current in the turbine elements and wetness effect on their characteristics. Problems of wetness separation and moving blade erosion as well as other turbine elements are extracted in a special section. Given are structural schemes of different methods of innerchannel and periphery wet removal as well as experimental materials on their effectiveness. Given are descriptions and critical analysis of a great number of typical constructions of nuclear power plant steam turbines, produced by native plants as well as by the main foreign firms. Considered also are constructions of outside separators and steam superheaters. Separately given is the problem of rotation frequency choise of nuclear power plant wet steam turbines. Represented are materials on turbine installation tests, considered are the problems of turbine starting and manoeuvrability, analyzed are their typical jailures and damages. One of the sections of the book is devoted to gas turbine installations of nuclear power plants. Different material on this theme scattered before in various sources is summarized in the book

  15. Collection and evaluation of complete and partial losses of off-site power at nuclear power plants

    International Nuclear Information System (INIS)

    Battle, R.E.

    1985-02-01

    Events involving loss of off-site power that have occurred at nuclear power plants through 1983 are described and categorized as complete or partial losses. The events were identified as plant-centered or grid-related failures. In addition, the causes of the failures were classified as weather, human error, design error, or hardware failure. The plant-centered failures were usually of shorter duration than the weather-related grid failures. For this reason, the weather-related events were reviewed in detail. Design features that may be important factors affecting off-site power system reliability were tabulated for most of the operating nuclear power plants. The tabulated information was provided to NRC for a statistical analysis to determine the importance of these design features for losses of off-site power. The frequency of losses of off-site power versus duration was estimated for three time periods. The frequency of loss of off-site power was estimated to be 0.09/reactor-year based on industry-wide data for the years 1959 through 1983

  16. Integrated Plant Safety Assessment: Systematic Evaluation Program. Millstone Nuclear Power Station, Unit 1, Northeast Nuclear Energy Company, Docket No. 50-245. Final report

    International Nuclear Information System (INIS)

    1983-02-01

    This report documents the review of the Millstone Nuclear Power Station, Unit 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit 1, is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license. This report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in November 1982

  17. Periodical safety review of the Goesgen-Daeniken nuclear power plant. Summary, results and evaluation; Periodische Sicherheitsueberpruefung fuer das Kernkraftwerk Goesgen-Daeniken. Zusammenfassung, Ergebnisse und Bewertung

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-15

    The Goesgen nuclear power plant (KKG) received its operational licence on September 9, 1978. The operational start-up of the plant went on into the year 1979, but there was a short interruption because of the accident in the Three Mile Island reactor on March 28, 1979. In May 1985 KKG submitted a request for raising the thermal reactor power from the then 2808 MW to 3002 MW. Based on the examination by the Federal Agency for the Safety of Nuclear Installations (HSK), the Swiss Federal Council granted the licence in two steps: in December 1985 for raising the thermal power to 2900 MW, and, in April 1992, to 3002 MW. The licence for the second step was given under the condition that some more experience was to be gained concerning the fuel rod cladding under higher loading. As part of the yearly re-licensing on restart after fuel assembly reloading, HSK confirmed that the plant status conformed to the legal requirements. In November 1986, HSK asked all Swiss nuclear power plant managers to state their opinions on proposed measures concerning severe accidents. Some of the measures were already in discussion; the Chernobyl accident on April 26, 1986, accelerated their implementation and was also a reason for the introduction of the measures against severe accidents. In this context, KKG carried out a risk study which led to the installation of a filtered pressure release system for the containment. Another consequence of the Chernobyl accident was the introduction of technical Periodical Safety Reviews (PSR) for all operating nuclear power plants. Central points of the PSR are: a) comparison with the continuously improving state-of-the-art of science and technology concerning safety precautions; b) a systematic evaluation of operating experience and plant status; c) the taking into account of probabilistic safety analyses in the overall evaluation of the plant. Within the framework of the examination of the overall plant, HSK also checks how its requirements concerning

  18. QA programs in nuclear power plants

    International Nuclear Information System (INIS)

    Ellingson, A.C.

    1976-01-01

    As an overview of quality assurance programs in nuclear power plants, the energy picture as it appears today is reviewed. Nuclear power plants and their operations are described and an attempt is made to place in proper perspective the alleged ''threats'' inherent in nuclear power. Finally, the quality assurance programs being used in the nuclear industry are described

  19. Investigation of the thermodynamic conditions in an underground nuclear power plant after core melt, in combination with a risk-oriented evaluation

    International Nuclear Information System (INIS)

    Hassmann, K.; Eyink, J.; Fischer, M.; Schwarzott, W.; Fabian, H.; Reineke, H.H.

    1983-01-01

    Based on the present state-of-the-art on the project ''Core Melt'', funded by the Federal Ministry of Research and Technology (BMFT), the thermodynamic conditions occurring in the safety containment (and/or) outer annulus of undergroud NPPs have been evaluated for selected release categories. In order to evaluate the pressure- and temperature history as well as the leakages, a coupled two-room-model was constructed. The most serious consequences following an accident in an underground nuclear power plant must be expected for the failure of containment isolation valve, as corresponding to release category 2 of the German Risk Study (DRS) which differs only slightly from the accident scenario gained by above-ground siting of nuclear power plants. (orig.) [de

  20. Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 4

    International Nuclear Information System (INIS)

    1985-03-01

    This report supplements the Safety Evaluation Report, NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), and Supplement No. 3 (January 1985) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report

  1. Safety Evaluation Report related to the operation of Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441). Supplement No. 6

    International Nuclear Information System (INIS)

    1985-04-01

    Supplement No. 6 to the Safety Evaluation Report (NUREG-0887) on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO), as applicants and owners, for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-540 and 50-441), has been prepared by the Office of the Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio, approximately 35 miles northeast of Cleveland, Ohio. This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 through 5 to that report

  2. Safety Evaluation Report related to the operation of Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441). Supplement No. 8

    International Nuclear Information System (INIS)

    1986-01-01

    Supplement No. 8 to the Safety Evaluation Report (NUREG-0887) on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO), as applicants and owners, for a license to operate the Perry Nuclear Power Plant, Units and 1 and 2 (Docket Nos. 50-440 and 50-441), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio, approximately 35 miles northeast of Cleveland, Ohio. This supplement reports the status of certain issues that has not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 through 7 to that report

  3. Safety Evaluation Report related to the operation of Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441). Supplement No. 7

    International Nuclear Information System (INIS)

    1985-11-01

    Supplement No. 7 to the Safety Evaluation Report (NUREG-0887) on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO), as applicants and owners, for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket No. 50-440 and 50-441) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio, approximately 35 miles northeast of Cleveland, Ohio. This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 through 6 to that report

  4. Safety evaluation report related to the operation of Perry Nuclear Power Plant, Units 1 and 2: Docket Nos. 50-440 and 50-441

    International Nuclear Information System (INIS)

    1983-01-01

    Supplement No. 2 to the Safety Evaluation Report on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group (CAPCO)), as applicants and owners, for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio. This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement No. 1 to that report

  5. Safety evaluation report related to the operation of Perry Nuclear Power Plant, Units 1 and 2. Docket Nos. 50-440 and 50-441

    International Nuclear Information System (INIS)

    1983-04-01

    Supplement No. 3 to the Safety Evaluation Report on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO), as applicants and owners, for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440/441), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio. This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 and 2 to that report

  6. Safety Evaluation Report related to the operation of Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441). Supplement No. 5

    International Nuclear Information System (INIS)

    1985-02-01

    Supplement No. 5 to the Safety Evaluation Report (NUREG-0887) on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, The Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO), as applicants and owners, for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio, approximately 35 miles northeast of Cleveland, Ohio. This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 through 4 to that report

  7. Safety evaluation report related to the operation of Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441). Supplement No. 10

    International Nuclear Information System (INIS)

    1986-09-01

    Supplement No. 10 to the Safety Evaluation Report (NUREG-0887) on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO), as applicants and owners for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio, approximately 35 miles northeast of Cleveland, Ohio. This supplement reports the status of certain issues and action items that had not been resolved or completed at the time of publication of the Safety Evaluation Report and Supplements Nos. 1 through 9 to that report

  8. Safety Evaluation Report related to the operation of Perry Nuclear Power Plant, Units 1 and 2, (Docket Nos. 50-440 and 50-441)

    International Nuclear Information System (INIS)

    1984-02-01

    Supplement No. 4 to the Safety Evaluation Report on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO), as applicants and owners, for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio. This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1, 2 and 3 to that report

  9. Nuclear security - New challenge to the safety of nuclear power plants

    International Nuclear Information System (INIS)

    Li Ganjie

    2008-01-01

    preparation and drills of emergency preparedness of nuclear facilities and carried out actively the preparation of anti terrorism in nuclear sector. Although the international community has been working on the nuclear security with a series of measures, according to the author there is a need author to discuss the following key perspectives. (1) It is essential to determine the definition of the nuclear security for power plants and come to a common understanding in the nuclear sector as soon as possible. (2) An international unified design basis threat to nuclear security of nuclear power plants should be developed to apply to the design of the nuclear security system of newly-built nuclear power plants and to evaluate the existing nuclear security system in the operating nuclear power plants so as to take improved measures. (3) The dividing of responsibilities between national government and nuclear power plants should be redefined in the new regime of nuclear security of nuclear power plants. (4) The relationship between the requirements of nuclear security and of the economy of nuclear power development should be balanced. (5) The technical standard system that suitable for new regime of nuclear security of nuclear power plants should be developed and improved to accelerate the enhancing of capability in nuclear security of nuclear power plants. It was concluded that nuclear terrorism is the common enemy to all the human beings. To strengthen the capacity of nuclear security of power plants, to ensure nuclear safety, are in the common interest and the responsibility of the entire international society. Recognizing the significance of strengthening the international cooperation on nuclear security, it is expected that the international society should closely cooperate together to establish the regime for nuclear security, share information and crack down nuclear terrorism. It was stated that China, as a responsible member of the international community, will continue to

  10. Applying Human-performance Models to Designing and Evaluating Nuclear Power Plants: Review Guidance and Technical Basis

    International Nuclear Information System (INIS)

    O'Hara, J.M.

    2009-01-01

    Human performance models (HPMs) are simulations of human behavior with which we can predict human performance. Designers use them to support their human factors engineering (HFE) programs for a wide range of complex systems, including commercial nuclear power plants. Applicants to U.S. Nuclear Regulatory Commission (NRC) can use HPMs for design certifications, operating licenses, and license amendments. In the context of nuclear-plant safety, it is important to assure that HPMs are verified and validated, and their usage is consistent with their intended purpose. Using HPMs improperly may generate misleading or incorrect information, entailing safety concerns. The objective of this research was to develop guidance to support the NRC staff's reviews of an applicant's use of HPMs in an HFE program. The guidance is divided into three topical areas: (1) HPM Verification, (2) HPM Validation, and (3) User Interface Verification. Following this guidance will help ensure the benefits of HPMs are achieved in a technically sound, defensible manner. During the course of developing this guidance, I identified several issues that could not be addressed; they also are discussed.

  11. Applying Human-performance Models to Designing and Evaluating Nuclear Power Plants: Review Guidance and Technical Basis

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, J.M.

    2009-11-30

    Human performance models (HPMs) are simulations of human behavior with which we can predict human performance. Designers use them to support their human factors engineering (HFE) programs for a wide range of complex systems, including commercial nuclear power plants. Applicants to U.S. Nuclear Regulatory Commission (NRC) can use HPMs for design certifications, operating licenses, and license amendments. In the context of nuclear-plant safety, it is important to assure that HPMs are verified and validated, and their usage is consistent with their intended purpose. Using HPMs improperly may generate misleading or incorrect information, entailing safety concerns. The objective of this research was to develop guidance to support the NRC staff's reviews of an applicant's use of HPMs in an HFE program. The guidance is divided into three topical areas: (1) HPM Verification, (2) HPM Validation, and (3) User Interface Verification. Following this guidance will help ensure the benefits of HPMs are achieved in a technically sound, defensible manner. During the course of developing this guidance, I identified several issues that could not be addressed; they also are discussed.

  12. Safety Evaluation Report related to the operation of Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441). Supplement No. 9

    International Nuclear Information System (INIS)

    1986-03-01

    Supplement No. 9 to the Safety Evaluation Report (NUREG-0887) on the application filed by the Cleveland Electric Illuminating Company on behalf of itself and as agent for the Duquesne Light Company, the Ohio Edison Company, the Pennsylvania Power Company, and the Toledo Edison Company (the Central Area Power Coordination Group or CAPCO) for a license to operate the Perry Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-440 and 50-441), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Lake County, Ohio, approximately 35 miles northeast of Cleveland, Ohio. This supplement reports the staff's evaluation findings pertaining to the earthquake event that occurred in the vicinity of the Perry Nuclear Power Plant site on January 31, 1986, and is limited to that evaluation. Future supplemental reports will continue reporting on the status of new or unresolved issues since Supplement No. 8 was issued in January 1986

  13. Exploiting nuclear plants in time

    International Nuclear Information System (INIS)

    Tran, Lionel

    2011-02-01

    This document outlines that the French fleet of 58 reactors is only 25 year old in average, and that nuclear safety is strongly regulated, and notably relies on improved indicators and on a decennial re-assessment. It outlines that nuclear energy is a response to energy challenges and that it is therefore relevant to operate the nuclear fleet beyond the initially foreseen lifetime (40 years). Due to maintenance and renewal activities, plants are supposed to be safer and more efficient. To guarantee an always safer and more efficient operation in time, five actions are highlighted: decennial controls, installation and equipment modifications, control and anticipation of installation and equipment wear, competencies and ability renewal, better knowledge of techniques and technologies

  14. Nuclear power plants and environment

    International Nuclear Information System (INIS)

    Agudo, E.G.; Penteado Filho, A.C.

    1980-01-01

    The question of nuclear power plants is analysed in details. The fundamental principles of reactors are described as well as the problems of safety involved with the reactor operation and the quantity and type of radioactive released to the environment. It shows that the amount of radioactive is very long. The reactor accidents has occurred, as three mile island, are also analysed. (M.I.A.)

  15. Operation of nuclear power plants

    International Nuclear Information System (INIS)

    Severa, P.

    1988-04-01

    The textbook for training nuclear power plant personnel is centred on the most important aspects of operating modes of WWER-440 reactors. Attention is devoted to the steady state operation of the unit, shutdown, overhaul with refuelling, physical and power start-up. Also given are the regulations of shift operation and the duties of individual categories of personnel during the shift and during the change of shifts. (Z.M.). 3 figs., 1 tab

  16. Nuclear power plant life management

    International Nuclear Information System (INIS)

    Rorive, P.; Berthe, J.; Lafaille, J.P.; Eussen, G.

    1998-01-01

    Several definitions can be given to the design life of a nuclear power plant just as they can be attributed to the design life of an industrial installation: the book-keeping life which is the duration of the provision for depreciation of the plant, the licensed life which corresponds to the duration for which the plant license has been granted and beyond which a new license should be granted by the safety authorities, the design life which corresponds to the duration specified for ageing and fatigue calculations in the design of some selected components during the plant design phase, the technical life which is the duration of effective technical operation and finally the economic life corresponding to the duration of profitable operation of the plant compared with other means of electricity production. Plant life management refers to the measures taken to cope with the combination of licensed, design, technical and economical life. They can include repairs and replacements of components which have arrived to the end of their life due to known degradation processes such as fatigue, embrittlement, corrosion, wear, erosion, thermal ageing. In all cases however, it is of great importance to plan the intervention so as to minimise the economic impact. Predictive maintenance is used together with in-service inspection programs to fulfil this goal. The paper will go over the methodologies adopted in Belgium in all aspects of electrical, mechanical and civil equipment for managing plant life. (author)

  17. Nuclear plant life cycle costs

    International Nuclear Information System (INIS)

    Durante, R.W.

    1994-01-01

    Life cycle costs of nuclear power plants in the United States are discussed. The author argues that these costs have been mishandled or neglected. Decommissioning costs have escalated, e.g. from $328 per unit in 1991 to $370 in 1993 for the Sacramento Municipal Utility District, though they still only amount to less than 0.1 cent per kWh. Waste management has been complicated in the U.S. by the decision to abandon civilian reprocessing; by the year 2000, roughly 30 U.S. nuclear power units will have filled their storage pools; dry storage has been delayed, and will be an expense not originally envisaged. Some examples of costs of major component replacement are provided. No single component has caused as much operational disruption and financial penalties as the steam generator. Operation and maintenance costs have increased steadily, and now amount to more than 70% of production costs. A strategic plan by the Nuclear Power Oversight Committee (of U.S. utilities) will ensure that the ability to correctly operate and maintain a nuclear power plant is built into the original design. 6 figs

  18. Regulatory framework for nuclear power plant operation

    International Nuclear Information System (INIS)

    Perez Alcaniz, T.; Esteban Barriendos, M.

    1995-01-01

    As the framework of standards and requirements covering each phase of nuclear power plant project and operation developed, plant owners defined their licensing commitments (codes, rules and design requirements) during the project and construction phase before start-up and incorporated regulatory requirements imposed by the regulatory Body during the licensing process prior to operation. This produces a regulatory framework for operating a plant. It includes the Licensing Basis, which is the starting point for analyzing and incorporating new requirements, and for re-evaluation of existing ones. This presentation focuses on the problems of applying this regulatory framework to new operating activities, in particular to new projects, analyzing new requirements, and reconsidering existing ones. Clearly establishing a plant's licensing basis allows all organizations involved in plant operation to apply the requirements in a more rational way. (Author)

  19. Occupational dose control in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Viktorsson, C.; Lochard, J.; Benedittini, M.; Baum, J.; Khan, T.A.

    1990-01-01

    Reduction in occupational exposure at nuclear power plants is desirable not only in the interest of the health and safety of plant personnel, but also because it enhances the safety and reliability of the plants. This report summarises the current trends of doses to workers at nuclear power plants and the achievements and developments regarding methods for their reduction

  20. Evaluation of environmental data relating to selected nuclear power plant sites. A synthesis and summary with recommendations

    International Nuclear Information System (INIS)

    Murarka, I.P.; Policastro, A.J.; Ferrante, J.G.; Daniels, E.W.; Marmer, G.J.

    1976-11-01

    An analysis of the fuel data gathered during the monitoring programs at seven nuclear power plant sites show no significant ecological impacts. This conclusion is within the constraints of the quality of available data from these sites. Current monitoring programs are, however, not designed to meet the needs for statistical analysis and, consequently, the monitoring data are often ill-suited for modern statistical procedures. Recommendations are proposed for revising monitoring schemes so that more precise conclusions can be made from fewer field measurements

  1. Engineering evaluation of selective ion-exchange radioactive waste processing at Susquehanna Nuclear Power Plant: Final report

    International Nuclear Information System (INIS)

    Vance, J.N.

    1989-01-01

    This final report describes the work performed of an engineering feasibility evaluation of the use and benefits of a selective ion exchange treatment process in the Susquehanna radwaste system. The evaluation addressed operability and processing capability concerns, radiological impacts of operating in the radwaste discharge mode, required hardware modifications to the radwaste and plant make-up systems, impacts on plant water quality limits and impacts on higher waste classifications. An economic analysis is also reported showing the economic benefit of the use of selective ion exchange. 1 ref., 4 figs., 13 tabs

  2. Model-based system engineering to evaluate I and C systems human performance in nuclear power plants - 15207

    International Nuclear Information System (INIS)

    Alves, P.; Mesquita, F.; Kessel, D.; Jung, J.

    2015-01-01

    In a Nuclear Power Plant, many advantages can be obtained by introducing digital instrumentation and control (I and C) systems, such as safety improvement in fault tolerance, self-testing, signal validation, better calibration and greater data capacity. The digital system must be able to meet the needs of the plants and also be compatible with operator capabilities, in a matter that will reduce human error probability. The safety systems must be reliable, with high functionality and availability, and it is essential to have redundancy, independence and diversity among components. I and C systems require a full understanding from plant operators. This paper is intent to propose, by using the C4ISR framework, a simpler and more comprehensive way to represent the architecture of the I and C system framework. The C4ISR framework gives the possibility to define the system under 3 different views: Operational View, Systems View, and Technical View; By using the CORE model-based system engineering software, the systems view of the Plant Protection System is analyzed

  3. Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391)

    International Nuclear Information System (INIS)

    Tam, P.S.

    1992-10-01

    This report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), Supplement No. 8 (January 1992), and Supplement No. 9 (June 1992) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units I and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER

  4. Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391)

    International Nuclear Information System (INIS)

    Tam, P.S.

    1991-04-01

    This report supplements the Safety Evaluation Report (SER), NUREG- 0847 (June 1982), Supplement No. 1' (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), and Supplement No. 5 (November 1990) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50--390 and 50--391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER

  5. Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391)

    International Nuclear Information System (INIS)

    Tam, P.S.

    1991-09-01

    This report supplements the Safety Evaluation Report (SER), NUREG- 0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), and Supplement No. 6 (April 1991) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER

  6. Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1983-12-01

    Supplement 20 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Unit 1 and Unit 2 (Docket Nos. 50-275 and 50-323), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the verification effort for Diablo Canyon Unit 1 that was performed between November 1981 and the present in response to Commission Order CLI-81-30 and an NRC letter of November 19, 1981 to the licensee. Specifically, Supplement 20 addresses those issues and other matters identified in Supplements 18 and 19 that must be resolved prior to Unit 1 achieving criticality and operating at power levels up to 5% of rated full power. This SER Supplement applies only to Diablo Canyon Unit 1

  7. Maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    Migaud, D.; Hutin, J.P.; Jouette, I.; Eymond, P.; Devie, P.; Cudelou, C.; Magnier, S.; Frydman, M.

    2016-01-01

    This document gathers different articles concerning the maintenance of the French nuclear power plants. The first article analyses the impact of the recent law on the energetic transition that sets the share of nuclear power at 50% of the electricity produced by 2025. A consequence may be the decommissioning of 17 to 20 reactors by 2025 and the huge maintenance program called 'Grand Carenage' whose aim is to extend operating life over 40 years will have to be re-considered in order to avoid useless expenses. The second article shows that in 2015 the French nuclear reactor fleet got very good results in terms of availability and safety. There were 49 scheduled outages and among them some ended ahead of time. The third article describes the specificities of the maintenance of a nuclear power plant, for instance the redundancy of some systems implies that maintenance has to deal with systems that have never functioned but must be ready to operate at any moment. Another specificity is the complexity of a nuclear power plant that implies an essential phase of preparation for maintenance operations. Because of safety requirements any maintenance operation has to be controlled, checked and may provide feedback. The fourth article presents the 'Grand Carenage' maintenance program that involves the following operations: the replacement of steam generators, the re-tubing of condensers, the replacement of the filtering drums used for cooling water, the testing of the reactor building, the hydraulic test of the primary circuit and the inspection of the reactor vessel. The fifth article focuses on the organization of the work-site for maintenance operations and the example of the Belleville-sur-Loire is described in the sixth article. Important maintenance operations like 'Grand Carenage' requires a strong collaboration with a network of specialized enterprises and as no reactor (except Flamanville EPR) is being built in France, maintenance

  8. Nuclear Plant Integrated Outage Management

    International Nuclear Information System (INIS)

    Gerstberger, C. R.; Coulehan, R. J.; Tench, W. A.

    1992-01-01

    This paper is a discussion of an emerging concept for improving nuclear plant outage performance - integrated outage management. The paper begins with an explanation of what the concept encompasses, including a scope definition of the service and descriptions of the organization structure, various team functions, and vendor/customer relationships. The evolvement of traditional base scope services to the integrated outage concept is addressed and includes discussions on changing customer needs, shared risks, and a partnership approach to outages. Experiences with concept implementation from a single service in 1984 to the current volume of integrated outage management presented in this paper. We at Westinghouse believe that the operators of nuclear power plants will continue to be aggressively challenged in the next decade to improve the operating and financial performance of their units. More and more customers in the U. S. are looking towards integrated outage as the way to meet these challenges of the 1990s, an arrangement that is best implemented through a long-term partnering with a single-source supplier of high quality nuclear and turbine generator outage services. This availability, and other important parameters

  9. Nuclear Power Plants in the World

    International Nuclear Information System (INIS)

    2003-01-01

    The Japan Atomic Industrial Forum (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2002 was made up on bases of answers on questionnaires from 65 electric power companies and other nuclear organizations in 28 countries and regions around the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities; and so forth. (J.P.N.)

  10. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    2008-01-01

    The Japan Atomic Industrial Forum, Inc. (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2007/2008 was made up on bases of answers on questionnaires from electric power companies and other nuclear organizations around the world by JAIF. This report is comprised of 18 items, and contains generating capacity of the plants; effect of the Niigata-ken chuetsu-oki earthquake; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities, and so forth. (J.P.N.)

  11. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    2007-01-01

    The Japan Atomic Industrial Forum, Inc. (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2005 was made up on bases of answers on questionnaires from electric power companies and other nuclear organizations around the world by JAIF. This report is comprised of 18 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities, and so forth. (J.P.N.)

  12. Nuclear Power Plants in the World

    International Nuclear Information System (INIS)

    2004-01-01

    The Japan Atomic Industrial Forum, Inc. (JAIF) used every year to summarize a trend survey on the private nuclear power plants in the world in a shape of the 'Nuclear power plants in the world'. In this report, some data at the end of 2003 was made up on bases of answers on questionnaires from 81 electric power companies and other nuclear organizations in 33 countries and regions around the world by JAIF. This report is comprised of 19 items, and contains generating capacity of the plants; current status of Japan; trends of generating capacity of operating the plants, the plant orders and generating capacity of the plants; world nuclear capacity by reactor type; status of MOX use in the world; location of the plants; the plants in the world; directory of the plants; nuclear fuel cycle facilities; and so forth. (J.P.N.)

  13. Role and assignments of a consultant for the evaluation and the construction of a nuclear power plant application to a project in Turkey

    International Nuclear Information System (INIS)

    Naigeon, J.; Menez, A.; Valentin, A.; Milliot, B.; Kutukcuoglu, A.; Sahin, S.

    1977-01-01

    For certain countries, the decision to build the first nuclear power plant raises serious problems, generally through lack of a sufficient number of specialists and of adequate experience to overcome the difficulties involved. The role of a consultant consists in making his knowledge and experience available to the future power plant owner, thus enabling him to make adequate and timely decisions and to build and operate the facility under optimum conditions. The consultant's assignments are determined case by case. Generally speaking, they concern the following aspects: specific site investigations; preparation of the invitations to bid and bid evaluation; preparation of the final project; technical assistance throughout the manufacture of the equipment, the construction and the erection stages and also during the plant acceptance and start-up tests; and safety analysis and participation in the preparation of the corresponding reports. To these assignments are added, where necessary, complementary studies such as: investigation of the problems involved in fuel procurement, management, storage and (or) reprocessing; cost estimation and financial planning; and definition of operating staff training requirements. At the outset, the future owner may also request advice as to the type of plant to adopt. All these aspects are analyzed in the light of the experience acquired during the performance of a contract, presently underway for the preparation of an invitation to bid for the construction of the first nuclear power plant (600 MWe) to be ordered by a national electricity utility [fr

  14. Simulators for nuclear power plants

    International Nuclear Information System (INIS)

    Ancarani, A.; Zanobetti, D.

    1983-01-01

    The different types of simulator for nuclear power plants depend on the kind of programme and the degree of representation to be achieved, which in turn determines the functions to duplicate. Different degrees correspond to different simulators and hence to different choices in the functions. Training of nuclear power plant operators takes advantage of the contribution of simulators of various degrees of complexity and fidelity. Reduced scope simulators are best for understanding basic phenomena; replica simulators are best used for formal qualification and requalification of personnel, while modular mini simulators of single parts of a plant are best for replay and assessment of malfunctions. Another category consists of simulators for the development of assistance during operation, with the inclusion of disturbance and alarm analysis. The only existing standard on simulators is, at present, the one adopted in the United States. This is too stringent and is never complied with by present simulators. A description of possible advantages of a European standard is therefore offered: it rests on methods of measurement of basic simulator characteristics such as fidelity in values and time. (author)

  15. Evaluation code for the dose due to the discharges of liquid effluents of the Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Lopez, Fabio O.; Boutet, Luis I.; Bruno, Hector A.; Gavini, Ricardo M.

    2004-01-01

    A new methodology is presented to assess the evaluation of the radiological impact to the population, due to the discharges to the environment of liquids effluents of Central Nuclear Embalse (CNE), located in the Province of Cordoba (Argentina). In order to carry out the dose evaluation, a code denominated EDDELIQ was developed, in this code the calculation of the radionuclides concentration in the water lake is made by means of a simple physical model of the type of complete mixture. The physical model is solved numerically by means of Runge Kutta method of second order. (author)

  16. Safety goals for nuclear power plant operation

    International Nuclear Information System (INIS)

    1983-05-01

    This report presents and discusses the Nuclear Regulatory Commission's, Policy Statement on Safety Goals for the Operation of Nuclear Power Plants. The safety goals have been formulated in terms of qualitative goals and quantitative design objectives. The qualitative goals state that the risk to any individual member of the public from nuclear power plant operation should not be a significant contributor to that individual's risk of accidental death or injury and that the societal risks should be comparable to or less than those of viable competing technologies. The quantitative design objectives state that the average risks to individual and the societal risks of nuclear power plant operation should not exceed 0.1% of certain other risks to which members of the US population are exposed. A subsidiary quantitative design objective is established for the frequency of large-scale core melt. The significance of the goals and objectives, their bases and rationale, and the plan to evaluate the goals are provided. In addition, public comments on the 1982 proposed policy statement and responses to a series of questions that accompanied the 1982 statement are summarized

  17. Licensing of nuclear power plants

    International Nuclear Information System (INIS)

    Witt, S. de.

    1984-01-01

    De Witt, who as an advocate represents the interests of citizen's initiatives in lawsuits concerning nuclear power plants, contests an independent scope of discretion of the administration and criticizes the reduction of the control density of courts which can be noticed recently. The shift of emphasis from the courts to the administration would not only change the balance within the constitutional separation of powers but could also lead to a weakening of the protection of fundamental rights. When assessing the technical risks he wants above all to take into account the possible extent of the damage. De Witt also shows that - beyond the discretion of refusal according to section 7 of the Atomic Energy Act and the exertion of influence of the executive on public utilities - it could be possible to develop means - not used until now - to control the social impacts of nuclear energy. (orig./HSCH) [de

  18. Development of a procedure for qualitative and quantitative evaluation of human factors as a part of probabilistic safety assessments of nuclear power plants. Part A

    International Nuclear Information System (INIS)

    Richei, A.

    1998-01-01

    The objective of this project is the development of a procedure for the qualitative and quantitative evaluation of human factors in the probabilistic safety assessment for nuclear power plants. The Human Error Rate Assessment and Optimizing System (HEROS) is introduced. The evaluation of a task with HEROS is realized in the three evaluation levels, i.e. 'Management Structure', 'Working Environment' and 'Man-Machine-Interface'. The developed expert system uses the fuzzy set theory for an assessment. For the evaluation of cognitive tasks evaluation criteria are derived also. The validation of the procedure is based on three examples, reflecting the common practice of probabilistic safety assessments and including problems, which cannot, respectively - only insufficiently - be evaluated with the established human risk analysis procedures. HERO applications give plausible and comprehensible results. (orig.) [de

  19. Evaluation of radiocesium concentrations in new leaves of wild plants two years after the Fukushima Dai-ichi Nuclear Power Plant accident

    International Nuclear Information System (INIS)

    Sugiura, Yuki; Shibata, Michihiro; Ogata, Yoshimune; Ozawa, Hajime; Kanasashi, Tsutomu; Takenaka, Chisato

    2016-01-01

    Radiocesium ( 137 Cs) transfer to plants immediately after the Fukushima Dai-ichi Nuclear Power Plant accident was investigated by collecting newly emerged leaf and soil samples between May 2011 and November 2012 from 20 sites in the Fukushima prefecture. Radiocesium concentrations in leaf and soil samples were measured to calculate concentration ratios (CR). Woody plants exhibited high CR values because 137 Cs deposited on stems and/or leaves were transferred to newly emerging tissues. The CR values in 2012 declined as compared to that in 2011. Exchangeable 137 Cs rates in soil (extraction rate) samples were measured at five sites. These rates decreased at four sites in 2012 and depended on environmental conditions and soil type. Both CR values and extraction rates decreased in 2012. However, CR values reflected the changes in extraction rates and characteristics of each species. Amaranthaceae, Chenopodiaceae, and Polygonaceae, which had been identified as Cs accumulators, presented no clear 137 Cs accumulation ability. In 2012, the perennial plant Houttuynia cordata and deciduous trees Chengiopanax sciadophylloides and Acer crataegifolium displayed high CR values, indicating that these species are 137 Cs accumulators and may be considered as potential species for phytoremediation. - Highlights: • Surface absorption causes higher 137 Cs in woody plants than in herbaceous plants. • Exchangeable 137 Cs rates decreased at four sites out of five in 2012. • Exchangeable 137 Cs rates and species-specific characteristics influence plant 137 Cs. • H. cordata, C. sciadophylloides, and A. crataegifolium displayed 137 Cs accumulation ability in 2012.

  20. Design of nuclear power plants

    International Nuclear Information System (INIS)

    Lobo, C.G.

    1987-01-01

    The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.) [pt

  1. Nuclear power plant operation 2016. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2017-05-15

    A report is given on the operating results achieved in 2016, events important to plant safety, special and relevant repair, and retrofit measures from nuclear power plants in Germany. Reports about nuclear power plants in Belgium, Finland, the Netherlands, Switzerland, and Spain will be published in a further issue.

  2. Environmental survey around EDF nuclear power plants

    International Nuclear Information System (INIS)

    Foulquier, L.

    1992-01-01

    Description of various types of environmental test carried out under the responsibility of the Operator of nuclear power plants in France, with taking Fessenheim nuclear power plant as an example: permanent monitoring of radioactivity, periodic radioecological assessments, main results of measurements taken, showing that there are no detectable effects of the plant on the environment, policy of openness by publication of these results

  3. Evaluation of radwaste minimization program of dry and wet active waste in the Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Luna-Garza, Hector

    2001-01-01

    A growing rate of radwaste volume production combined with an increase of both, costs and associated dose involved in its treatment and disposition processes have created a serious problem to the Laguna Verde Nuclear Power Plant (BWR, two Units, 682 Mwe each) in Mexico. Due the lack of a Final Repository in the country, the solution in the short or long terms relies on the success of a continuous and aggressive minimization program mainly based on modifications and upgrades applied to these processes. Technical and administrative strategies adopted by LVNPP for the reduction of Liquid Effluents and Dry and Wet Active Waste in the next three years are described. Based on the results of the LVNPP current radwaste process systems, an estimated accumulation of 11,502 m 3 by the year 2035 will exceed the actual on-site storage capacity. If the strategies succeed, this production would fall to an expected manageable volume of 4067 m 3 . (author)

  4. An Evaluation of the Physical Environments of a Nuclear Power Plants for Human Factors Review in Periodic Safety Review

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Yong Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    Currently, operation of a nuclear power plants(NPP) is highly emphasized by the integrity of the H/W and the human factors security, so the periodic safety review(PSR) is performed to NPP. The PSR activities on human factors include physical environments (illumination, noise, vibration, temperature and humidity etc). The review on these physical environments is to verify the possible affect to the human error during the operation of the man machine interface. Physical environments affect the health, job stress and job satisfaction of NPP's employees. On the ground of the reason, we need integrating the management program for the sufficient satisfaction of the regulatory basis and standards of physical environment and considering a health, a job stress and satisfaction of NPP's employees. So, this paper describes the planning of the setup procedures of physical environments and the adequate management program for the field applications in NPPs.

  5. An Evaluation of the Physical Environments of a Nuclear Power Plants for Human Factors Review in Periodic Safety Review

    International Nuclear Information System (INIS)

    Kim, Dae Ho; Lee, Yong Hee

    2006-01-01

    Currently, operation of a nuclear power plants(NPP) is highly emphasized by the integrity of the H/W and the human factors security, so the periodic safety review(PSR) is performed to NPP. The PSR activities on human factors include physical environments (illumination, noise, vibration, temperature and humidity etc). The review on these physical environments is to verify the possible affect to the human error during the operation of the man machine interface. Physical environments affect the health, job stress and job satisfaction of NPP's employees. On the ground of the reason, we need integrating the management program for the sufficient satisfaction of the regulatory basis and standards of physical environment and considering a health, a job stress and satisfaction of NPP's employees. So, this paper describes the planning of the setup procedures of physical environments and the adequate management program for the field applications in NPPs

  6. Nuclear Plant Aging Research (NPAR) program plan

    International Nuclear Information System (INIS)

    1991-06-01

    A comprehensive Nuclear Plant Aging Research (NPAR) Program was implemented by the US NRC office of Nuclear Regulatory Research in 1985 to identify and resolve technical safety issues related to the aging of systems, structures, and components in operating nuclear power plants. This is Revision 2 to the Nuclear Plant Aging Research Program Plant. This planes defines the goals of the program the current status of research, and summarizes utilization of the research results in the regulatory process. The plan also describes major milestones and schedules for coordinating research within the agency and with organizations and institutions outside the agency, both domestic and foreign. Currently the NPAR Program comprises seven major areas: (1) hardware-oriented engineering research involving components and structures; (2) system-oriented aging interaction studies; (3) development of technical bases for license renewal rulemaking; (4) determining risk significance of aging phenomena; (5) development of technical bases for resolving generic safety issues; (6) recommendations for field inspection and maintenance addressing aging concerns; (7) and residual lifetime evaluations of major LWR components and structures. The NPAR technical database comprises approximately 100 NUREG/CR reports by June 1991, plus numerous published papers and proceedings that offer regulators and industry important insights to aging characteristics and aging management of safety-related equipment. Regulatory applications include revisions to and development of regulatory guides and technical specifications; support to resolve generic safety issues; development of codes and standards; evaluation of diagnostic techniques; (e.g., for cables and valves); and technical support for development of the license renewal rule. 80 refs., 25 figs., 10 tabs

  7. Evaluating nuclear communications

    International Nuclear Information System (INIS)

    Middlemiss, Nigel

    1998-01-01

    plant, and how messages on nuclear safety, economics, waste and exports were accepted and rejected by the media; - how large the payback can be from proactive - not just reactive - media work, as evidenced in an analysis for an industrial group involved in North Sea oil plant decommissioning. A demonstration will be given showing the parameters that can be measured by media analysis and presented through charts, tables and other graphical means: totals of media 'impressions', favourability ratings for companies and industries, leading issues, leading publications or TV/radio programmes, leading sources of information or opinion quoted in the media, successful and unsuccessful corporate messages, and leading 'bylines'. (author)

  8. Commercialization of nuclear power plant decommissioning technology

    International Nuclear Information System (INIS)

    Williams, D.H.

    1983-01-01

    The commercialization of nuclear power plant decommissioning is presented as a step in the commercialization of nuclear energy. Opportunities for technology application advances are identified. Utility planning needs are presented

  9. Basic safety principles for nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Shiguan

    1989-01-01

    To ensure the safety operation of nuclear power plant, one should strictly adhere to the implelmentation of safety codes and the establishment of nuclear safety code system, as well as the applicable basic safety principles of nuclear power plants. This article briefly introduce the importance of nuclear codes and its economic benefits and the implementation of basic safety principles to be accumulated in practice for many years by various countries

  10. TOSHIBA CAE system for nuclear power plant

    International Nuclear Information System (INIS)

    Machiba, Hiroshi; Sasaki, Norio

    1990-01-01

    TOSHIBA aims to secure safety, increase reliability and improve efficiency through the engineering for nuclear power plant using Computer Aided Engineering (CAE). TOSHIBA CAE system for nuclear power plant consists of numbers of sub-systems which had been integrated centering around the Nuclear Power Plant Engineering Data Base (PDBMS) and covers all stage of engineering for nuclear power plant from project management, design, manufacturing, construction to operating plant service and preventive maintenance as it were 'Plant Life-Cycle CAE System'. In recent years, TOSHIBA has been devoting to extend the system for integrated intelligent CAE system with state-of-the-art computer technologies such as computer graphics and artificial intelligence. This paper shows the outline of CAE system for nuclear power plant in TOSHIBA. (author)

  11. Application of analysis technology in nuclear plant

    International Nuclear Information System (INIS)

    Takaoka, Keiko; Miura, Hiromi; Umeda, Kenji

    1996-01-01

    Recently, thanks to the rapid improvement of EWS performance, the authors have been able to carry out design evaluation comparatively, easily, utilizing computational fluid dynamics (CFD) technology. The Nuclear Plant Engineering Department has carried out some analyses in the past several years with the main purpose of evaluating the design of nuclear reactor internals. These studies included ''Thermal Hydraulic Analysis for Top Plenum'' and ''Flow Analysis for Lower Plenum''. It is considered to be a special matter in thermal hydraulic analysis of the top plenum that temperature distribution has been estimated with a relatively small number of meshes by means of an imaginary spray nozzle, and in the flow analysis for the lower plenum that flow distribution has been found to change largely, depending on the reactor internals. One of the ways to confirm the safety of nuclear plants, detailed structural analysis, is required for all possible combinations of transient and load conditions during operation. In particular, it is very important to clarify the thermal stress behavior under operating conditions and to evaluate fatigue analysis in accordance with the Code Requirements. However, it is very complicated and it takes a lot of time. A new system was developed which can operate continuously all of the definitions of the analytical model, the analyzation of pressurized thermal and external stress, and editing reports. In this paper, the authors introduce this system and apply it to a pressurized water reactor

  12. Nuclear Plant Modification in a Risk-Informed Environment

    International Nuclear Information System (INIS)

    Gallucci, Raymond H.V.

    2002-01-01

    This paper examines a specific nuclear power plant modification performed in a risk-informed regulatory environment. It quantifies both the permanent and temporary effects of the modification, and performs a cost-benefit evaluation. (authors)

  13. Safety aspects of nuclear power plant ageing

    International Nuclear Information System (INIS)

    1990-01-01

    The nuclear community is facing new challenges as commercial nuclear power plants (NPPs) of the first generation get older. At present, some of the plants are approaching or have even exceeded the end of their nominal design life. Experience with fossil fired power plants and in other industries shows that reliability of NPP components, and consequently general plant safety and reliability, may decline in the middle and later years of plant life. Thus, the task of maintaining operational safety and reliability during the entire plant life and especially, in its later years, is of growing importance. Recognizing the potential impact of ageing on plant safety, the IAEA convened a Working Group in 1985 to draft a report to stimulate relevant activities in the Member States. This report provided the basis for the preparation of the present document, which included a review in 1986 by a Technical Committee and the incorporation of relevant results presented at the 1987 IAEA Symposium on the Safety Aspects of the Ageing and Maintenance of NPPs and in available literature. The purpose of the present document is to increase awareness and understanding of the potential impact of ageing on plant safety; of ageing processes; and of the approach and actions needed to manage the ageing of NPP components effectively. Despite the continuing growth in knowledge on the subject during the preparation of this report it nevertheless contains much that will be of interest to a wide technical and managerial audience. Furthermore, more specific technical publications on the evaluation and management of NPP ageing and service life are being developed under the Agency's programme, which is based on the recommendations of its 1988 Advisory Group on NPP ageing. Refs, figs and tabs

  14. Qualification of nuclear power plant operations personnel

    International Nuclear Information System (INIS)

    1984-01-01

    With the ultimate aim of reducing the possibility of human error in nuclear power plant operations, the Guidebook discusses the organizational aspects, the staffing requirements, the educational systems and qualifications, the competence requirements, the ways to establish, preserve and verify competence, the specific aspects of personnel management and training for nuclear power plant operations, and finally the particular situations and difficulties to be overcome by utilities starting their first nuclear power plant. An important aspect presented in the Guidebook is the experience in training and qualification of nuclear power plant personnel in various countries: Argentina, Belgium, Canada, Czechoslovakia, France, Federal Republic of Germany, Spain, Sweden, United Kingdom and United States of America

  15. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

  16. Large nuclear steam turbine plants

    International Nuclear Information System (INIS)

    Urushidani, Haruo; Moriya, Shin-ichi; Tsuji, Kunio; Fujita, Isao; Ebata, Sakae; Nagai, Yoji.

    1986-01-01

    The technical development of the large capacity steam turbines for ABWR plants was partially completed, and that in progress is expected to be completed soon. In this report, the outline of those new technologies is described. As the technologies for increasing the capacity and heightening the efficiency, 52 in long blades and moisture separating heaters are explained. Besides, in the large bore butterfly valves developed for making the layout compact, the effect of thermal efficiency rise due to the reduction of pressure loss can be expected. As the new technology on the system side, the simplification of the turbine system and the effect of heightening the thermal efficiency by high pressure and low pressure drain pumping-up method based on the recent improvement of feed water quality are discussed. As for nuclear steam turbines, the actual records of performance of 1100 MW class, the largest output at present, have been obtained, and as a next large capacity machine, the development of a steam turbine of 1300 MWe class for an ABWR plant is in progress. It can be expected that by the introduction of those new technologies, the plants having high economical efficiency are realized. (Kako, I.)

  17. Overall quality assurance program requirements for nuclear power plants

    International Nuclear Information System (INIS)

    1992-09-01

    This standard contains the requirements for the owner's overall quality assurance program for a nuclear power plant. This program encompasses all phases of a nuclear power plant life cycle, including site evaluation, design, procurement, manufacturing, construction and installation, commissioning, operation, and decommissioning. It covers the activities associated with specifying, directing, and administering the work to be done during these phases, and the evaluation and integrated of the activities and programs of participants

  18. New nuclear plant design and licensing process

    International Nuclear Information System (INIS)

    Luangdilok, W.

    1996-01-01

    This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

  19. Nuclear Power Plant (NPP) safety in Brazil

    International Nuclear Information System (INIS)

    Lederman, L.

    1980-01-01

    The multidisciplinary aspects of the activities involved in the nuclear power plant (NPP) licensing, are presented. The activities of CNEN's technical staff in the licensing of Angra-1 and Angra-2 power plants are shown. (E.G.) [pt

  20. Preliminary regulatory assessment of nuclear power plants vulnerabilities

    International Nuclear Information System (INIS)

    Kostadinov, V.; Petelin, S.

    2004-01-01

    Preliminary attempts to develop models for nuclear regulatory vulnerability assessment of nuclear power plants are presented. Development of the philosophy and computer tools could be new and important insight for management of nuclear operators and nuclear regulatory bodies who face difficult questions about how to assess the vulnerability of nuclear power plants and other nuclear facilities to external and internal threats. In the situation where different and hidden threat sources are dispersed throughout the world, the assessment of security and safe operation of nuclear power plants is very important. Capability to evaluate plant vulnerability to different kinds of threats, like human and natural occurrences and terrorist attacks and preparation of emergency response plans and estimation of costs are of vital importance for assurance of national security. On the basis of such vital insights, nuclear operators and nuclear regulatory bodies could plan and optimise changes in oversight procedures, organisations, equipment, hardware and software to reduce risks taking into account security and safety of nuclear power plants operation, budget, manpower, and other limitations. Initial qualitative estimations of adapted assessments for nuclear applications are shortly presented. (author)