WorldWideScience

Sample records for backfitting

  1. Backfitting of research reactors

    International Nuclear Information System (INIS)

    Delrue, R.; Noesen, T.

    1985-01-01

    The backfitting of research reactors covers a variety of activities. 1. Instrumentation and control: Control systems have developed rapidly and many reactor operators wish to replace obsolete equipment by new systems. 2. Pool liners: Some pools are lined internally with ceramic tiles. These may become pervious with time necessitating replacement, e.g. by a new stainless steel liner. 3. Heat removal system: Deficiencies can occur in one or more of the cooling system components. Upgrading may require modifications of the system such as addition of primary loops, introduction of deactivation tanks, pump replacement. Recent experience in such work has shown that renewal, backfitting and upgrading of an existing reactor is economically attractive since the related costs and delivery times are substantially lower than those required to install a new research reactor

  2. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  3. Case study on the use of PSA methods: Backfitting decisions

    International Nuclear Information System (INIS)

    1991-04-01

    This case study illustrates the process of using probabilistic risk assessment (PRA) method to evaluate proposed backfits of nuclear power plants (NPP), which are intended to enhance the plant safety by improving equipment operability. Some examples of situations in which PRA techniques have been used to address backfit issues at operating NPPs are summarized. 2 refs, 5 figs, 4 tabs

  4. Licensing requirements for backfit incinerators at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Dodge, R.L.; Edwards, C.W.; Wilson, B.

    1984-01-01

    This paper, and the project it reports on, examines the licensing requirements for backfit incinerators at operating power plants. Analysis was made of incinerating low-level dry radioactive waste in a backfit incinerator at an existing power plant. The operation of the incinerator has been studied from viewpoints of operator safety, consequence of system failures including worst case scenarios, and radiological impact for normal and upset conditions. Analysis showed that releases under all normal operating or upset conditions are an extremely small fraction of the applicable limits. Nuclear Regulatory Commission review concluded that the document produced as a result of this project was useful as a design guide and of value in licensing backfit incinerators. 1 table

  5. Backfitting in Rossendorf research reactor control and instrumentation system

    International Nuclear Information System (INIS)

    Klebau, J.; Seidler, S.

    1985-01-01

    The paper generally describes a decentralized Hierarchical Information System (HIS) which has been developed for backfitting in Rossendorf Research Reactor (RFR) control and instrumentation system. The RFR was put into operation in 1957 and reconstructed from 2 MW up to a thermal power of 10 MW at the end of the sixties. Backfitting is planned by use of an advanced computerized control system for the next years. Main tasks of HIS are: Processmonitoring, online-disturbance analysis, technical diagnosis, direct digital control and use of a special industrial robot for discharging of irradiated materials out of the reactor. Experiences obtained by HIS during a testperiod will be presented. (author)

  6. Modification and backfitting in safety related systems at Ringhals 2

    International Nuclear Information System (INIS)

    Lidh, B.; Stroemqvist, E.

    1995-08-01

    This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Ringhals-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 4 refs

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  8. Use of PSA for design modifications and backfitting

    International Nuclear Information System (INIS)

    Evans, M.G.K.

    1997-01-01

    The aim of this presentation is to gain an understanding of how the living PSA can be used to evaluate proposed design changes and backfitting modifications. The topics are: Process for evaluation of design change. Application to various types of change - Design changes; Procedure changes; Temporary plant modifications; Design deviations. 1 fig

  9. Limited probabilistic risk assessment applications in plant backfitting

    International Nuclear Information System (INIS)

    Desaedeleer, G.

    1987-01-01

    Plant backfitting programs are defined on the basis of deterministic (e.g. Systematic Evaluation Program) or probabilistic (e.g. Probabilistic Risk Assessment) approaches. Each approach provides valuable assets in defining the program and has its own advantages and disadvantages. Ideally one should combine the strong points of each approach. This chapter summarizes actual experience gained from combinations of deterministic and probabilistic approaches to define and implement PWR backfitting programs. Such combinations relate to limited applications of probabilistic techniques and are illustrated for upgrading fluid systems. These evaluations allow sound and rational optimization systems upgrade. However, the boundaries of the reliability analysis need to be clearly defined and system reliability may have to go beyond classical boundaries (e.g. identification of weak links in support systems). Also the implementation of upgrade on a system per system basis is not necessarily cost-effective. (author)

  10. Seismic hazard assessment in intra-plate areas and backfitting

    International Nuclear Information System (INIS)

    Asmis, G.J.K.; Eng, P.

    2001-01-01

    Typically, fuel cycle facilities have been constructed over a 40 year time period incorporating various ages of seismic design provisions ranging from no specific seismic requirements to the life safety provisions normally incorporated in national building codes through to the latest seismic nuclear codes that provide not only for structural robustness but also include operational requirements for continued operation of essential safety functions. The task is to ensure uniform seismic risk in all facilities. Since the majority of the fuel cycle infrastructure has been built the emphasis is on re-evaluation and backfitting. The wide range of facilities included in the fuel cycle and the vastly varying hazard to safety, health and the environment suggest a performance based approach. This paper presents such an approach, placed in an intra-plate setting of a Stable Continental Region (SCR) typical to that found in Eastern Canada. (author)

  11. A proposed approach to backfit decision-making using risk assessment and benefit-cost methodology

    International Nuclear Information System (INIS)

    O'Donnell, E.P.; Raney, T.J.

    1984-01-01

    This paper outlines a proposed approach to backfit decision-making which utilizes quantitative risk assessment techniques, benefit-cost methodology and decision criteria. In general terms, it is structured to provide an objective framework for decision-making aimed at ensuring a positive return on backfit investment while allowing for inclusion of subjective value judgments by the decision-maker. The distributions of the independent variables are combined to arrive at an overall probability distribution for the benefit-cost ratio. In this way, the decision-maker can explicitly establish the probability or level of confidence that a particular backfit will yield benefits in excess of cost. An example is presented demonstrating the application of methodology to a specific plant backfit. (orig.)

  12. NPP Evaluation, backfitting and life extension. An engineering viewpoint

    International Nuclear Information System (INIS)

    Gonzalez Lopez, A.

    1993-01-01

    During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

  13. PSA based plant modifications and back-fits

    International Nuclear Information System (INIS)

    1997-01-01

    The mandate of Principal Working Group No. 5 - Risk Assessment states that 'The group should deal with the technology and methods for identifying contributors to risk and assessing their importance, and appropriate exchanges of information on current research'. Since being formulated in 1982, along with this mandate, the group has also endeavored to develop a common understanding of the different approaches taken in risk assessment. The focus of this report is to provide knowledge to experts on the role Probabilistic Safety Assessment (PSA) has had in safety decision making. PSA is a powerful tool for improving Nuclear Power Plant safety by identifying weaknesses in design or operation and setting priorities for plant modifications and back-fits. While the use is well recognised, it is also true that any safety decision is generally based on several elements, both probabilistic and deterministic. This document provides a general overview of insights gained from the representative set of examples collected from Member countries (Finland, France, Germany, Japan, Korea, Netherlands, Spain, Sweden, Switzerland, United Kingdom, United States). The report starts with basic types of plant modifications which were carried out (e.g. hardware or software, important or minor, etc.) and the characteristics of the PSAs used in the examples (e.g. level and scope, specific or generic, on-going or terminated, etc.). The insights gained from this small collection are then reviewed. The appendix gives a full text version of the Member country contributions

  14. Review of instrumentation and control system backfitting in Loviisa nuclear power station in Finland

    International Nuclear Information System (INIS)

    Ekman, I.

    1985-01-01

    Loviisa nuclear power station consists of two 440 MWe PWR units. Loviisa 1, the first nuclear power plant in Finland, started commercial operation in 1977 and Loviisa 2 in 1981. In general the instrumentation and control systems of the plant have performed well and the general design solutions have been satisfactory. In spite of this after the start-up of the plant a lot of backfitting work of instrumentation and control systems has been carried out. Major contribution of the backfitting work is due to the fact that safety requirements become stricter as new technological innovations find application and as lessons are learnt from experience with operating plants. Especially the lessons learnt from TMI-2 have influenced Loviisa instrumentation and control system backfitting. Only a minor part of the backfitting work has been the changing of components that have turned out to be unreliable or correcting design deficiencies. The paper gives an overview of the backfitting work of the instrumentation and control systems of Loviisa plant and some examples of adopted solutions are described in more detail. New measurements for accident monitoring are described. A description is given concerning changes that have been made or that are planned to the plant protection system. The environmental qualification of safety related equipment located inside containment has been assessed. Work in the field of man-machine communication is discussed. On-line surveillance system of plant main components is described. (author)

  15. Modification and backfitting at the Oskarshamn Nuclear Power Plant Unit 2 in safety related systems

    International Nuclear Information System (INIS)

    Karlsson, Leif; Nilsson, Ove; Lidh, B.

    1995-05-01

    This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Oskarshamn-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 3 refs

  16. Examination of the conditions governing nuclear licensing, supervision, and backfitting of installations

    International Nuclear Information System (INIS)

    Papier, H.J.

    1991-01-01

    The expert opinion examines the following subjects: the licensing requirements and conditions as defined in the Atomic Energy Act, the Radiation Protection Ordinance and the X-ray Ordinance, their relation to each other and differences that would need a systematic coordination; licensing requirements and the term 'nuclear installation', the discretionary powers of a licensing authority; licensing and other official authorizations and questions arising therefrom, as e.g. concentrating effects; problems in connection with plan approval; the procedure of licensing by stages; prevention of damage by a dynamic system of precautionary measures and the backfitting of nuclear installations; licensing in connection with backfitting and the problem of a participation of the public in the licensing procedure; decommissioning of nuclear installations. (HSCH) [de

  17. Modification and backfitting at the Barsebaeck Nuclear Power Plant Unit 1 and 2 in safety related systems

    International Nuclear Information System (INIS)

    Karlsson, Leif; Nilsson, Ove; Lidh, B.

    1995-05-01

    This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Barsebaeck, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 3 refs

  18. Radiation protection during backfitting or dismantling work in the controlled area of nuclear facilities

    International Nuclear Information System (INIS)

    Baumann, J.; Kausch, S.; Palmowski, J.

    1980-01-01

    Backfitting measures or dismantling activities within the controlled area put special requirements on radiological protection. This is to be shown by the example of the following cases. Sanitation of the general decontamination services of the Karlsruhe Nuclear Research Center; waste water, equipment decontamination, incineration and packaging facility; dismantling and disposal of high-radiation components including decontamination of buildings of the Eurochemic reprocessing plant at Mol; reconstruction of the HDR plant for safety experiments together with waste management for components and systems, as e.g. pressure vessel internals, pipes etc.; exchange of the steam dryer and the water separator including planning of the conditioning process in the Wuergassen nuclear power plant. This lecture deals with the engineering and organizational problems, especially accounting for radiological protection and enters into planning of measures for radiological protection, their organization and execution, problems of direct and remote-controlled work also being discussed. The question of personnel qualification is also commented on. (orig.) [de

  19. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  20. I and C related aspects during backfitting of a special heat removal system (UNS) for a BWR at Brunsbuettel

    International Nuclear Information System (INIS)

    Fasko, P.

    1985-01-01

    The BWR at Brunsbuettel (KKB, 770 MWe), north of the Federal Republic of Germany (FRG), went into commercial operation in 1976. In 1976 the Bundesminister des Inneren (BMI) of the FRG (federal responsibility for superior safety aspects of NPP's) asked for the implementation of a special emergency heat removal system (Unabhaengiges Notstandssystem -UNS) for the NPP Brunsbuettel (KKB). The goal of this backfitting is to cope with events which were not postulated in the original design of the plant and, to further reduce the residual risk. After completion of the detailed planning and the corresponding safety assessment, the authorities granted the construction and operation license for the UNS beginning November 1982. Site construction of the new buildings began just afterwards

  1. Backfit modifications to operating radwaste systems

    International Nuclear Information System (INIS)

    Giorgione, D.M.; Dresser, C.D.; Irving, T.J.; La Marca, J.T.

    1984-01-01

    A comprehensive radwaste modification project to replace corroded tanks and piping and increase liquid radwaste storage capacity is described. The major factor potentially affecting both schedule and cost is the low labor productivity associated with work in radiation areas. Engineering design and construction planning activities were formulated to minimize the impact on system operation and control exposure during construction. A detailed Health Physics Plan was developed which provides for decontamination of work areas consistent with ALARA/cost benefit considerations

  2. Advances in safety analysis and backfitting design of piping systems

    International Nuclear Information System (INIS)

    Bartholome, G.; Habip, L.M.

    1993-01-01

    Major topics during a safety evaluation of pipework in operating nuclear power stations are external events (e.g. earthquakes) and internal events (e.g. postulated pipe ruptures). Some of the corresponding material and structural mechanics aspects of the integrity of such systems are reviewed. This includes leak-before-break considerations and nonlinear response under strong base excitation or due to simulated breaks and valve closure. (author)

  3. Backfitting of the nuclear plant V1 power control system

    International Nuclear Information System (INIS)

    Karpeta, C.; Rubek, J.; Stirsky, P.

    1985-01-01

    The paper deals with some aspects of implementation of modifications into the Czechoslovak nuclear plant V1 control system as called for on the basis of experience gained during the first period of the plant operation. Brief description of the plant power control system and its main functions is given. Some deficiencies in the system performance during abnormal conditions are outlined and measures taken to overcome them are presented. (author)

  4. Operating cost reduction by optimization of I and C backfitting strategy

    International Nuclear Information System (INIS)

    Kraft, Heinz-U.

    2002-01-01

    Full text: The safe and economic operation of a nuclear power plant requires a large scope of automation systems to act properly in combination. The associated maintenance costs, necessary to test these systems periodically and to repair or to replace them partly or completely, are one important factor in the overall operating costs of a nuclear power plant. Reducing these costs by reducing the maintenance effort could decrease the availability of the power plant and by this way increase the operating costs significantly. The minimization of the overall operating costs requires a well-balanced maintenance strategy taking into account all these opposite influences. The replacement of an existing I and C system by a new one reduces the maintenance cost in the long term and increases the plant availability. However, it requires some investments in the short term. On the other hand the repair of an I and C system avoids investments, but it doesn't solve the aging problems. That means maintenance costs will increase in the long term and the plant availability could be decreased. An optimized maintenance strategy can be elaborated on a plant specific base taking into account the residual lifetime of the plant, the properties of the installed I and C systems as well as their influence on the plant availability. As a general result of such an optimization performed by FANP it has been found as a rule that the replacement of I and C systems becomes the most economic way the longer the expected lifetime is and the stronger the I and C system influences, the availability of the plant. (author)

  5. Regulatory and backfit analysis: Unresolved safety issue A-45, shutdown decay heat removal requirements

    International Nuclear Information System (INIS)

    1988-11-01

    All light water reactors require decay heat to be removed subsequent to reactor shutdown. Interruption of the decay heat removal function could lead to severe consequences. Concerns about the reliability of the systems and components that assist in the decay heat removal process and the potentially severe consequences of a complete loss of decay heat removal resulted in establishing the requirements for decay heat removal as an unresolved safety issue (USI) designated USI A-45, ''Shutdown Decay Heat Removal Requirements.'' This report presents the regulatory analysis for USI A-45. It includes (1) a summary of the issue, (2) the proposed technical resolution, (3) alternative resolutions considered by the Nuclear Regulatory Commission, (4) an assessment of the benefits and costs of all alternatives considered, and (5) the decision rationale. 23 refs., 9 figs., 39 tabs

  6. Backfitting of East European nuclear power plants to be financed by German funds

    International Nuclear Information System (INIS)

    1992-01-01

    The countries which were represented at the World Economic Summit in Munich have offered support to the new independent countries of the former Soviet Union as well as of Middle and Eastern Europe within the framework of a multi-national campaign to increase the safety of Soviet-made nuclear power plants. This campaign plans short- and long-term measures. The short-term measures should be financed via bilateral assistance by the countries and supplemented by a multilateral fund. For these measures a total amount of an estimated 700 million dollars will be required. (orig./HSCH) [de

  7. Methodological approach for the seismic backfitting of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Galli, P.; Muzzi, F.; Ruggieri, G.; Zola, M.

    1993-01-01

    In the frame of the assessment of the seismic adequacy of the operating Nuclear Power Plants in East Europe, the main problem to match with is the difficulty to work about already existing plants. Moreover consolidated standards and procedures for seismic design, verification and qualification exist for new structures and equipment, then the extension to operating plants requires a lot of engineering judgement. The paper highlights the importance of: identification of seismic safety related systems and components; site specific seismic input definition in agreement with international standards; computation of seismic loads accounting for soil-structure interaction and appropriate structural modelling; overall stability verification of the plant (soil bearing capacity, soil liquefaction, sliding, overturning); ductility effects in evaluation of seismic protection; engineering process for the qualification of components and systems and walkdown procedures and identification of remedial measures (easy fixes and complex fixes). Some examples are reported referred to the more recent ISMES activities in the field

  8. A digital, decentralized power station control system with bus-transmission facilitates the problem of backfitting

    International Nuclear Information System (INIS)

    Kaiser, G.E.; Schemmel, R.R.

    1985-01-01

    Current NPP control equipment technology is essentially characterized by the transmission of information in parallel using individual cables, and utilizes hardwired techniques for the processing of information. Progress in the area of semiconductor development characterized by micro-processors and LSI-circuits, has opened up new possibilities for the solution of the control tasks. The new power station control system PROCONTROL P utilizes these possibilities

  9. Compilation of backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1989-01-01

    The efforts for the modernization of the FRG-reactors within the last two years and at present are: Measures against water leakage through the concrete and along beam tubes, repair of both cooling towers, modernization of the ventilation system, measures for fire protection, activities in water chemistry and water quality, installation of a double tubing for parts of the primary piping of the FRG-1, replacement of instrumentation, process control system (operation and monitoring system) and alarm system, installation of a cold neutron source, enrichment reduction for the FRG-1. Planned activities are: Renewal of the emergency power supply, installation for internal lightning protection, compressed air system. (orig.) With 26 figs., 1 tab [de

  10. Report of a consultants meeting on backfittings and safety enhancement measures in NPPs with WWER 440/213 reactors. Extrabudgetary programme on the safety of WWER NPPS

    International Nuclear Information System (INIS)

    1994-01-01

    The purpose of this Consultants' Meeting held by the IAEA in Vienna from 11-15 April 1994 within the framework of the Extrabudgetary Programme on WWER Safety was to review and analyze safety issues revealed during operation and through analyses of NPPs with WWER 440/213 reactors. The initial list of safety issues based on the available reports from various studies had been prepared by the IAEA secretariat before the meeting, together with indications of safety enhancement measures proposed in various NPP units. During the meeting, the underlying safety concerns and actual technical status of the plants were discussed and the ranking of the safety issues was considered. 58 refs, 1 tab

  11. Nuclear Power Plant Control and Instrumentation activities in Argentina during 1987-1989

    International Nuclear Information System (INIS)

    Lorenzetti, J.R.

    1990-01-01

    The paper describes the status of the NPP Control and Instrumentation in Argentina. The following fields are very briefly described: Dynamic analysis, plant simulator, plant backfitting, in service inspection, robotics and tools, man-machine communication. (author)

  12. Safety analysis of the Morsleben radioactive waste repository (ERAM)

    International Nuclear Information System (INIS)

    Beise, E.; Biesold, H.; Gruendler, D.; Handge, P.; Lange, F.; Larue, J.; Mielke, H.; Mueller, W.; Peiffer, F.; Pfeffer, W.; Wurtinger, W.; Jaritz, W.; Meister, D.; Schnier, H.

    1991-03-01

    Stocktaking of the present ERAM situation and the safety assessment show that there are no hazards which would require a stop of operation at the moment. However, backfitting measures have been identified, part of which has to be taken without delay, such as underground fire protection. Those backfitting measures do not depend on the operational state of the plant, and can therefore be implemented during operation. (orig.) [de

  13. Future control room design (modernization of control room systems)

    International Nuclear Information System (INIS)

    Reischl, Ludwig; Freitag, Timo; Dergel, Rene

    2009-01-01

    In the frame of lifetime extension for nuclear power plants the modernization of the complete safety and operational control technology will be digitalized. It is also recommended to modernize the operator facilities, monitoring systems in the control room, the back-up shut-down center and the local control stations. The authors summarize the reasons for the modernization recommendations and discuss possible solutions for display-oriented control rooms. A concept for control room backfitting includes generic requirements, requirements of the local authorities, ergonomic principles information content and information density, and the design process. The backfitting strategy should include a cooperation with the operational personnel, The quality assurance and training via simulator needs sufficient timing during the implementation of the backfitting.

  14. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  15. Experience with emergency diesels at the Swiss NPP Goesgen (KKG)

    International Nuclear Information System (INIS)

    Steffen, W.

    1986-01-01

    The Goesgen nuclear power plant, a 970 MWe KWU pressurized water reactor, is fitted with 4 x 50 X emergency diesels and 2 x 100 % special emergency (Notstand) diesel units. Since the start-up tests of the diesels in 1977 several severe incidents occurred. As a consequence, different back-fitting actions were taken on the diesels and the emergency electrical System. The presentation will treat the following subjects: - lay-out of the onsite electrical power sources, - experiences and problems, - back-fitting measures, - periodic testing of the diesels. (author)

  16. Bibliography for Advancement Study.

    Science.gov (United States)

    1987-07-01

    PAVINSiT ’,JOO.6(; tandtarl )rgenization end ftgu.atkns of the hope and 10 YC’,AQC ardl iocal staff Itroctives ft) juidanco o~nI tfparliq4 Staff st~jdlos...Monitor Panel (DAMP), NAVSEA OD 54648 (POSEIDON), OD 55257 (TRIDENT BACKFIT) EIMB, General Handbook, NAVSEA SEQOO-00-EIM-100 -- Secs 2 and 4 Fluid Power

  17. New requirements, rules and regulations, and vested rights of nuclear power plants

    International Nuclear Information System (INIS)

    Raetzke, C.

    2006-01-01

    The article deals with the question whether new requirements can be imposed on existing nuclear power plants. It was promoted by the fact that the German Federal Ministry for the Environment currently is working on a thorough revision of German nuclear regulations. When looking at backfitting requirements, the all-important question is whether new findings show that the provisions taken in the license to guarantee the 'necessary precautions' (as defined in the German Atomic Energy Act) contain errors or omissions; only in this case can the authority demand that remedial measures, including backfitting, be taken. Beyond that, German nuclear law contains no obligation for operators to improve and develop safety still further. This applies regardless of whether new requirements are justified by new technical possibilities or new scientific analyses or whether they are prompted by a mere abstract re-evaluation of the safety level to be achieved. In the former case, if there are good technical or scientific reasons, the operators, as a rule, will perform backfitting voluntarily. Pursuant to these criteria, the article covers three categories of backfitting requirements and illustrates them by examples. These general principles are also valid when a new set of regulations - as planned by the BMU - are put into effect and applied. They may lead to existing plants not having to comply fully with the requirements contained in new regulations. (orig.)

  18. Use of PSA and PSC in the regulatory process in The Netherlands

    International Nuclear Information System (INIS)

    Versteeg, M.F.; Vos, D.

    1994-01-01

    The paper presents the regulatory requirements, thinking, and plans regarding the use of plant specific PSAs in the Netherlands, the actual use of probabilistic safety criteria (PSC) in the existing regulations and the PSA based plant modifications and backfits. 1 fig., 6 tabs

  19. Topform '92: the safe and reliable operation of LWR NPPs. Vol. I

    International Nuclear Information System (INIS)

    1993-01-01

    The proceedings contain 23 invited plenary session papers. All have been inputted to INIS. The topics covered include safety principles, management and organization, operational training and simulation, inspection, maintenance and component replacement, backfitting experience, instrumentation, man-machine interface, software and procedures. (Z.S.)

  20. Borssele: giving it a new lease of life, to 2007 and beyond

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    A major $250 million modification project is underway at the twin-loop Borssele PWR in the Netherlands. After the backfitting programme, which will raise safety levels to the current state of the art, this 20 year old plant will still be competitive with modern combined cycle gas-fired stations, according to the economic analysis that the Dutch have done. (author)

  1. Topform '92: the safe and reliable operation of LWR NPPs. Vol. II

    International Nuclear Information System (INIS)

    1993-01-01

    Out of the 54 poster papers contained in the proceedings, 53 were inputted to the INIS system. The topics covered include operational training and simulation, inspection, maintenance and component replacement, backfitting experience, instrumentation, man-machine interface, software and procedures. (Z.S.)

  2. Improved sealing for in-core systems

    International Nuclear Information System (INIS)

    Dunford, S.

    1989-01-01

    The in-core instrumentation sealing nozzles designed by Framatome have three mechanical seals in series instead of the one traditional seal, and are pressurized by simply tightening up the nozzle covers. They have been installed from the start on all Framatome PWRs, as well as having been backfitted on Belgium and Yugoslavian units and chosen for the Chinese Qinshan plant. (author)

  3. Nonparametric additive regression for repeatedly measured data

    KAUST Repository

    Carroll, R. J.

    2009-05-20

    We develop an easily computed smooth backfitting algorithm for additive model fitting in repeated measures problems. Our methodology easily copes with various settings, such as when some covariates are the same over repeated response measurements. We allow for a working covariance matrix for the regression errors, showing that our method is most efficient when the correct covariance matrix is used. The component functions achieve the known asymptotic variance lower bound for the scalar argument case. Smooth backfitting also leads directly to design-independent biases in the local linear case. Simulations show our estimator has smaller variance than the usual kernel estimator. This is also illustrated by an example from nutritional epidemiology. © 2009 Biometrika Trust.

  4. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    Adinolfi, R.; Noviello, L.

    1990-01-01

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

  5. Mixture of Regression Models with Single-Index

    OpenAIRE

    Xiang, Sijia; Yao, Weixin

    2016-01-01

    In this article, we propose a class of semiparametric mixture regression models with single-index. We argue that many recently proposed semiparametric/nonparametric mixture regression models can be considered special cases of the proposed model. However, unlike existing semiparametric mixture regression models, the new pro- posed model can easily incorporate multivariate predictors into the nonparametric components. Backfitting estimates and the corresponding algorithms have been proposed for...

  6. Control rooms in German nuclear power plants

    International Nuclear Information System (INIS)

    Hoffmann, E.

    1999-01-01

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

  7. Enhanced radiation weapons, radiation fields and shielding protection

    International Nuclear Information System (INIS)

    Hehn, G.

    1989-01-01

    Suitable and effective protection against neutron weapons can be well provided for in new residential or other buildings, and backfitting of existing buildings can be done at reasonable cost. Studies have shown that 100 per cent protection can be achieved by the fact found, that the neutron beams penetrate the walls in the corner between outer wall and room ceiling, and this pathway can be plugged by neutron absorbing material. (DG) [de

  8. Belgian national report

    International Nuclear Information System (INIS)

    Berthe, J.

    1995-01-01

    At last IWG-LMNPP meeting, the approach on nuclear power plant life management in Belgium was presented. The present report focuses on results of in-service monitoring of major equipment, specifically reactor internals, reactor top-head penetrations and steam generators. Status of major backfitting on steam generators and balance of plant is developed as well as developments in the field of thermal stratification and qualification of ultrasonic inspection methods and personnel for in-service inspection. (author). (Abstract only)

  9. Ergonomics: an aid to system design

    International Nuclear Information System (INIS)

    McCafferty, D.B.

    1990-01-01

    In recent years, the engineering community has recognized that ergonomics can make significant contributions to system design. Working together engineers and ergonomists can create designs that effectively meet system goals. By considering the role of humans and technology in the context of systems and by reducing the potential for errors, gains can be made in overall system reliability. Such efforts can reduce the need for costly backfits and increase system efficiency. (author)

  10. Steam generator replacement at Doel 3 NPP (Belgium)

    International Nuclear Information System (INIS)

    Danhier, B.

    1993-01-01

    The reasons are presented that led to the conclusion that the most cost-effective strategy for the Doel 3 unit was the immediate replacement of the SG. Discussed are the advantages and drawbacks of the replacement techniques, the so-called 2, 3 and 4 cuts methods. The advantages are emphasized of intensive use of computer aided engineering in this kind of backfitting. The methodology applied to combine a power uprating of 10% over the nominal power with the steam generator replacement is presented. (author) 1 fig

  11. 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

  12. Advances in Canadian regulatory practice

    International Nuclear Information System (INIS)

    Waddington, J.G.

    1993-03-01

    The new General Amendments to the Regulations, new recommendations on dose limits, developments in techniques and safety thinking, and aging of plant are all contributing to the need for a significant number of new regulatory document on a wide range of topics. this paper highlights a number of initiatives taken in response to these pressures, giving a brief background to the initiative and, where possible, outlining some of the ideas in the document licensing guides on new dose limits, dosimetry, safety analysis, reliability, fault tree analysis, reporting requirements, human factors, software, the ALARA principle, backfitting and the licensing process. (Author) 29 refs., fig., 4 tabs

  13. SIEMENS

    International Nuclear Information System (INIS)

    2001-01-01

    This CD is multimedia presentation of programme safety upgrading of Bohunice V1 NPP. This chapter contains information about Siemens and it participation on reconstruction of Bohunice V1 and V1 NPPs. It consists of next parts: (1) FRAMATOME ANP - worldwide activities of the FRAMATOME are presented; (2) Nuclear power engineering - present activities focus on: Upgrading and Backfitting (Siemens WWER activities since 1971); Electrical instrumentation and control systems; Fuel assemblies and related services; Reactor development and construction of new plants; (3) Safety improvement; (4) Siemens in Slovakia (activities of Siemens in Slovakia during 1993-2000 are presented); (5) More than 150-year history

  14. Upgrading safety of NPPs with RBMK-1000 reactors by implementation of the first priority measures and activities

    International Nuclear Information System (INIS)

    1996-01-01

    After the accident at the Chernobyl Unit 4 reactor, extensive debates were in place about the future of nuclear power industry, its safety and the role of nuclear power in human life. The major conclusion drawn from those discussions is that the energy demands and ecological problems could not be resolved without further development of nuclear industry. However, the continued development of nuclear power industry, first and foremost, should rest on a wide range of actions aimed at assuring the quality of design and construction of new NPPs, the quality of operation of the existing plants and by means of their backfitting. 1 ref., 3 figs, 1 tab

  15. Use of digital photography for power plant retrofits

    International Nuclear Information System (INIS)

    Kamba, J.J.

    1995-01-01

    One of the latest advancements in electronic tools for reducing engineering and drafting effort is the use of digital photography (DP) for retrofit and betterment projects at fossil and nuclear power plants. Sargent and Lundy (S and L) has effectively used digital photography for condition assessments, minor backfit repairs, thermo-lag fire wrap assessments and repairs, and other applications. Digital photography offers several benefits on these types of projects including eliminating the need for official repair drawings and providing station maintenance with a true 3-D visualization of the repair

  16. Combustion Engineering adjusts to slump in nuclear orders

    International Nuclear Information System (INIS)

    Masters, R.

    1980-01-01

    It is three years since Combustion Engineering (C-E) received an order for a nuclear steam system supplier and it could be three or four years before a new order is placed. Although C-E will not work through its current backlog until the late 1990s, the lack of new business and the needs for backfitting are having a major impact on the way the company operates. C-E's determination to stay in the nuclear business is as strong as ever. (author)

  17. Hydrogen countermeasures and activity retention by filtered venting for WWER-440/V230 NPP confinement

    International Nuclear Information System (INIS)

    Feuerbach, R.

    2001-01-01

    In order to prevent loss of confinement integrity caused by steam and hydrogen generation nuclear power plants in the Federal Republic of Germany as well as in most other European countries have been or will be back-fitted with a system for filtering the confinement atmosphere prior to release to the environment and a system for reducing and measuring the H 2 concentration inside the confinement. For these tasks systems for confinement atmosphere control are presented, capable of: handling high H 2 production rates and cleaning of high contaminated confinement atmosphere. (author)

  18. The development and application of quantitative methods in licensing nuclear power plants

    International Nuclear Information System (INIS)

    Cave, L.; Kastenberg, W.E.; Tweedy, J.N.

    1984-01-01

    The development and application of two quantitative methods, which could be used as part of the decision making process in nuclear power plant licensing, are decribed. These methods are the use of quantitative screening criteria to assess the adequacy of the safety functions in existing plants and the use of cost/benefit analysis to determine limits to the cost effective expenditure on ''back-fitting'' to improve safety. It is shown that the results obtained by the two methods are not necessarily compatible with one another. The need for clear guidance from regulatory bodies on the choice of some major parameters used in cost/benefit analysis is demonstrated. (orig.)

  19. Dealing with control rod guide tube support pin cracking in French PWRs

    International Nuclear Information System (INIS)

    Guicherd, L.

    1984-01-01

    Cracking and failure of control rod guide tube support pins has been encountered at a number of PWRs around the world. To deal with the problem, the French embarked on an extremely ambitious backfitting programme, involving the installation of replacement pins at all their operating 900MWe units. This highly successful programme, which will be completed in 1985, has been carried out with very low occupational doses and, in the last two years, has required no extensions to annual refuelling outage periods at the plants concerned. The French approach has involved a number of innovations, which should be of considerable interest to other PWR owners worldwide. (author)

  20. Nuclear power plant control and instrumentation in Switzerland

    International Nuclear Information System (INIS)

    Voumard, A.

    1992-01-01

    In Switzerland five NPPs are in operation and none is planned or is under construction. The three oldest NPPs are backfitted with an additional safety system. In the field of I and C, efforts are essentially directed to maintaining high performance and to improve the safety of the plants in operation. Three of these plants are about 20 years old and a significant part of their I and C equipment has to be replaced. This is an ongoing process which is carried out stage by stage mostly during the annual shutdown. Measures to avoid or mitigate severe accidents, including core melting, have been taken or are planned. (author). 1 tab

  1. Guidelines for control room systems design. Working material. Report

    International Nuclear Information System (INIS)

    1993-01-01

    This report contains comprehensive technical and methodological information and recommendations for the benefit of Member States for advice and assistance in ''NPP control room systems'' design backfitting existing nuclear power plants and design for future stations. The term ''Control Room Systems'' refers to the entire human/machine interface for the nuclear stations - including the main control room, back-ups control room and the emergency control rooms, local panels, technical support centres, operating staff, operating procedures, operating training programs, communications, etc. Refs, figs and tabs

  2. Control room systems design for nuclear power plants

    International Nuclear Information System (INIS)

    1995-07-01

    This publication provides a resource for those who are involved in researching, managing, conceptualizing, designing, manufacturing or backfitting power plant control room systems. It will also be useful to those responsible for performing reviews or evaluations of the design and facilities associated with existing power plant control room systems. The ultimate worth of the publication, however, will depend upon how well it can support its users. Readers are invited to provide comments and observations to the IAEA, Division of Nuclear Power. If appropriate, the report will subsequently be re-issued, taking such feedback into account. Refs, figs and tabs

  3. 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

  4. Systems required during and after an earthquake. Summary report. WWER-1000 nuclear power plants

    International Nuclear Information System (INIS)

    Monette, P.

    1995-01-01

    The scope of this document is to list the mechanical, instrumentation and electrical components required during and after earthquake, in order to achieve and maintain safe shutdown conditions of a WWER-1000 type nuclear power plant. The main objective pursued in establishing the systems and equipment list is to provide guidance for the design and implementation of the backfits which are necessary to increase seismic resistance of the components required after earthquake. The presented list is established on generic basis, i.e. it is applicable to any specific WWER-1000

  5. Upgrading the safety of VVER-440/V-230

    International Nuclear Information System (INIS)

    Kelm, P.; Wenk, W.

    1995-01-01

    Besides measures seeking to restore the status as laid down in the project design, especially backfitting measures must be mentioned which serve to ensure component and pipe integrity. Ensuring component integrity is a problem not only of RPV embrittlement, but also of failure prevention. This aspect was not always taken into account properly. Further activities in the field of component integrity will focus on backing the brittle fracture evaluation of the RPV; qualifying the leak-before-breack criterion for the main pipes and in areas with screwed connections; qualifying the program of in-service inspections. Several operators are currently in the process of drafting backfitting programs. The upgrading measures envisaged must be checked as to their balanced nature. In certain plants, the integrity of the RPV coud turn out to be the weak spot in upgrading measures. As a consquence, concepts seeking to achieve upgrading for long periods of time may differ from one location to the next and even between units. Extensive modifications in systems engineering and building structures generally must be evaluated against the expected improvement in safety of the whole plant. (orig.) [de

  6. Replacement strategy for obsolete plant computers

    International Nuclear Information System (INIS)

    Schaefer, J.P.

    1985-01-01

    The plant computers of the first generation of larger nuclear power plants are reaching the end of their useful life time with respect to the hardware. The software would be no reason for a system exchange but new tasks for the supervisory computer system, availability questions of maintenance personnel and spare parts and the demand for improved operating procedures for the computer users have stimulated the considerations on how to exchange a computer system in a nuclear power plant without extending plant outage times due to exchange works. In the Federal Republic of Germany the planning phase of such backfitting projects is well under way, some projects are about to be implemented. The base for these backfitting projects is a modular supervisory computer concept which has been designated for the new line of KWU PWR's. The main characteristic of this computer system is the splitting of the system into a data acquisition level and a data processing level. This principle allows an extension of the processing level or even repeated replacements of the processing computers. With the existing computer system still in operation the new system can be installed in a step-by-step procedure. As soon as the first of the redundant process computers of the data processing level is in operation and the data link to the data acquisition computers is established the old computer system can be taken out of service. Then the back-up processing computer can be commissioned to complete the new system. (author)

  7. Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  8. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  9. Technology, safety and costs of decommissioning a refernce boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

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

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  10. Investigations related to the management of wastes from the Rossendorf research reactor. Project study

    International Nuclear Information System (INIS)

    Quaas, H.; Boerner, H.J.; Wolf, D.

    1991-01-01

    The objective of the study consists in working out a technologically continuous waste management concept for the spent nuclear fuel from the Rossendorf research reactor (RFR) in operation since 1957. After backfitting, including replacement of the reactor pressure vessel, the facility went into operation again in January 1990 and continued until 30 June 1991. The fuel elements spent so far are stored compactly under water in two storage ponds near the RFR which have a total capacity of 3498 storage places. 2100 of them are already occupied by fuel elements which differ in their geometrical dimensions, in the type of fuel, and in number. During normal operation 200 fuel elements arise annually. In this regard, the Central Institute of Nuclear Research at Rossendorf faces the following problems: The USSR does not take back the spent nuclear fuel; about 70% of their storage capacity is already utilized; the storage ponds have to be backfitted to guarantee leak tightness. That requires a temporary relocation of the fuel elements and waste management start in January 1994. There are no waste management technologies and techniques available for the RFR, and the integrity of several fuel elements is endangered due to already visible corrosion effects. (orig.) [de

  11. Modernisation of the Borssele NPP reactor protection system

    International Nuclear Information System (INIS)

    Plas, Y. van der

    1993-01-01

    For several years the obligation to evaluate the safety level of nuclear power plants in The Netherlands against the state of the art has been required in the licenses of such plants. This was leading to backfitting programs for both nuclear power plants in The Netherlands. These programmes are in an engineering phase at present. One of the plants to be retrofitted is the Borssele NPP. This is a 450 MWe PWR, in operation since 1973. Design and construction is from Siemens/KWU. Its concept is from an earlier date than for instance the KWU-Konvoi design and thus shows more ramification and less separation in process, electrical and instrumental redundancies than more recent plant types. To combat dependencies in failure modes an additional bunkerized civil structure, Building 33, had been already erected in 1985. This building meets redefined requirements for flood, gas cloud explosion, and external fire. It provides space for the functionally and physically separated secondary loop. Independency has been acquired by its own power supply. These systems provide a certain back up for the primary volume control, the core injection system, and the auxiliary feed water supply in case of an external event. The present backfitting program is also utilizing the design considerations of these systems

  12. Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

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

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  13. Balancing safety and economics

    International Nuclear Information System (INIS)

    Kroeger, W.; Fischer, P.U.

    2000-01-01

    The safety requirements of NPPs have always aimed at limiting societal risks. This risk approach initially resulted in deterministic design criteria and concepts. In the 1980s the paradigm 'safety at all costs' arose and often led to questionable backfitting measures. Conflicts between new requirements, classical design concepts and operational demands were often ignored. The design requirements for advanced reactors ensure enhanced protection against severe accidents. Still, it is questionable whether the 'no-damage-outside-the-fence' criteria can be achieved deterministically and at competitive costs. Market deregulation and utility privatisation call for a balance between safety and costs, without jeopardising basic safety concepts. An ideal approach must be risk-based and imply modern PSAs and new methods for cost-benefit and ALARA analyses, embed nuclear risks in a wider risk spectrum, but also make benefits transparent within the context of a broader life experience. Governments should define basic requirements, minimum standards and consistent comparison criteria, and strengthen operator responsibility. Internationally sufficient and binding safety requirements must be established and nuclear technology transfer handled in a responsible way, while existing plants, with their continuous backfitting investments, should receive particular attention. (orig.)

  14. Evaluation of severe accident risks and the potential for risk reduction: Surry Power Station, Unit 1: Draft report for comment

    International Nuclear Information System (INIS)

    Benjamin, A.S.; Boyd, G.J.; Kunsman, D.M.; Murfin, W.B.; Williams, D.C.

    1987-02-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a particular pressurized water reactor with a subatmospheric containment (Surry, Unit 1). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally lower than previously evaluated in the Reactor Safety Study (RSS). However, certain unresolved issues (such as direct containment heating) caused the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. None of the postulated safety options appears to be cost effective for the Surry power plant. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150

  15. Evaluation of severe accident risks and the potential for risk reduction: Surry Power Station, Unit 1: Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, A.S.; Boyd, G.J.; Kunsman, D.M.; Murfin, W.B.; Williams, D.C.

    1987-02-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a particular pressurized water reactor with a subatmospheric containment (Surry, Unit 1). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally lower than previously evaluated in the Reactor Safety Study (RSS). However, certain unresolved issues (such as direct containment heating) caused the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. None of the postulated safety options appears to be cost effective for the Surry power plant. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150.

  16. Nuclear Criticality Safety Department Qualification Program

    International Nuclear Information System (INIS)

    Carroll, K.J.; Taylor, R.G.; Worley, C.A.

    1996-01-01

    The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of highly qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. This document defines the Qualification Program to address the NCSD technical and managerial qualification as required by the Y-1 2 Training Implementation Matrix (TIM). This Qualification Program is in compliance with DOE Order 5480.20A and applicable Lockheed Martin Energy Systems, Inc. (LMES) and Y-1 2 Plant procedures. It is implemented through a combination of WES plant-wide training courses and professional nuclear criticality safety training provided within the department. This document supersedes Y/DD-694, Revision 2, 2/27/96, Qualification Program, Nuclear Criticality Safety Department There are no backfit requirements associated with revisions to this document

  17. Gradient Plasticity Model and its Implementation into MARMOT

    Energy Technology Data Exchange (ETDEWEB)

    Barker, Erin I.; Li, Dongsheng; Zbib, Hussein M.; Sun, Xin

    2013-08-01

    The influence of strain gradient on deformation behavior of nuclear structural materials, such as boby centered cubic (bcc) iron alloys has been investigated. We have developed and implemented a dislocation based strain gradient crystal plasticity material model. A mesoscale crystal plasticity model for inelastic deformation of metallic material, bcc steel, has been developed and implemented numerically. Continuum Dislocation Dynamics (CDD) with a novel constitutive law based on dislocation density evolution mechanisms was developed to investigate the deformation behaviors of single crystals, as well as polycrystalline materials by coupling CDD and crystal plasticity (CP). The dislocation density evolution law in this model is mechanism-based, with parameters measured from experiments or simulated with lower-length scale models, not an empirical law with parameters back-fitted from the flow curves.

  18. Topical issues in nuclear, radiation and radioactive waste safety. Contributed papers

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA International Conference on Topical Issues in Nuclear, Radiation and Radioactive Waste Safety was held in Vienna, Austria, 30 August - 4 September 1998 with the objective to foster the exchange of information on topical issues in nuclear, radiation and radioactive waste safety, with the aim of consolidating an international consensus on: the present status of these issues; priorities for future work; and needs for strengthening international co-operation, including recommendations for the IAEA's future activities. The document includes 43 papers presented at the Conference dealing with the following topical issues: Safety Management; Backfitting, Upgrading and Modernization of NPPs; Regulatory Strategies; Occupational Radiation Protection: Trends and Developments; Situations of Chronic Exposure to Residual Radioactive Materials: Decommissioning and Rehabilitation and Reclamation of Land; Radiation Safety in the Far Future: The Issue of Long Term Waste Disposal. A separate abstract and indexing were provided for each paper

  19. Measures taken to improve nuclear safety on EdF PWRs in operation

    International Nuclear Information System (INIS)

    Kus, J.-P.; Norvez, G.

    1993-01-01

    In parallel with its major nuclear programme (56 PWR units in service or under construction), France has developed an original philosophy in the field of Nuclear Safety. This comprehensive philosophy ensures a fine balance and coordination between design and operation, it provides a methodology to design, construct and operate a safe nuclear plant. Actual experience is then continuously compared to the initial expectation and the methodology refined whenever necessary. This methodology is fully applied to the new 1400 MWe plant series presently under construction. The essential elements are also backfitted into all previous units, thereby giving them an equivalent level of safety. The French PWR park can therefore be considered as very homogeneous with regard to its safety level, regarding both its design and operation. (author)

  20. Procedures for maintenance and repairs

    International Nuclear Information System (INIS)

    Pickel, E.

    1981-01-01

    After a general review of the operation experience in the history of more than 12 operating years, the organization in the plant will be shown with special aspect to quality assurance, capacity of the workshops and connected groups as radiation protection, chemical laboratories etc. The number, time intervals and manpower effort for the repeating tests will be discussed. Reasons and examples for back-fitting activities in the plant are given. Besides special repair and maintenance procedures as repair of the steam generators, in-service inspection of the reactor pressure vessel, repair of a feed-water pipe and repair of the core structure in the pressure vessel, the general system to handle maintenance and repair-work in the KWO-plant will be shown. This includes also the detailed planning of the annual refueling and revision of the plant. (orig./RW)

  1. Analysis of public comments on the proposed rule on nuclear power plant license renewal

    International Nuclear Information System (INIS)

    1991-12-01

    This report provides a summary and analysis of public comments on the proposed license renewal rule for the nuclear power plants (10 CFR Part 54) published in the Federal Register on 17 July 1990. It also documents the NRC's resolution of the issues raised by the commenters. Comments from 121 organizations and 76 individuals were reviewed and analyzed to identify the issues, including those pertaining to the adequacy of the licensing basis, the performance of an integrated plant assessment, backfit considerations, and need for public hearings. The analysis included grouping of commenters' views according to the issues raised. The public comments analyzed in this report were taken into consideration in the development of the final rule and revisions to the supporting documents

  2. Reliability evaluation under new requirements for the safety system control technology in nuclear power plants. Analysis of the application practice; Zuverlaessigkeitsbewertung unter neuen Anforderungen an Sicherheitsleittechnik in Kernkraftwerken. Analysen der Anwendungspraxis

    Energy Technology Data Exchange (ETDEWEB)

    Jopen, Manuela; Quester, Claudia; Roemer, Sarah; Sommer, Dagmar; Stiller, Jan; Ulrich, Birte

    2016-10-15

    In many German nuclear power plants computer-based of programmable control systems are integrated in the operational control technology; but there exist significant differences. For a reliability evaluation the backfitting/retrofitting of safety relevant equipment has to be considered. The problems result from the fact that at the time of license application only few exactly formulated requirements for computer based of programmable control systems were available. The RSK considers the existing DIN norms as not sufficient, further the norms exceeding methods are supposed to be necessary to guarantee an optimal and verifiable design against common cause failures of control systems. A comparison of national and international norms has shown the existence of detailed requirements in several standards. The transferability of these requirements is discussed.

  3. Identifying measures to balance the risk profile of the Tihange 2 NPP

    International Nuclear Information System (INIS)

    D'Eer, A.M.; Monniez, J.J.

    2001-01-01

    In Belgium, each Nuclear Power Plant is subject to a periodic safety reassessment. In this context, it was found to be desirable to perform a Probabilistic Safety Assessment (PSA) in support of the ten yearly back-fitting process. The Tihange 2 NPP is a 3-loop PWR having a thermal capacity of 2905 MW. Analysis of the plant's risk profile shows that implementing feasible measures for improvement of the shutdown risk, would be beneficial. This is because a configuration leading to significant risk, namely cold pressurization when the residual heat removal system is lost during reduced primary inventory, thus can be avoided. As a result the risk between reactor shutdown and power operation will be balanced. The presentation describes the lessons learnt regarding the Tihange 2 shutdown PSA model and the expected benefits following implementation of one of the proposed measures. (author)

  4. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  5. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

    Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

  6. The economic impact of reactor transients

    International Nuclear Information System (INIS)

    Rossin, A.D.; Vine, G.L.

    1984-01-01

    This chapter discusses the cost estimation of transients and the causal relationship between transients and accidents. It is suggested that the calculation of the actual cost of a transient that has occurred is impossible without computerized records. Six months of operating experience reports, based on a survey of pressurized water reactors (PWRs) and boiling water reactors (BWRs) conducted by the Nuclear Safety Analysis Center (NSAC), are analyzed. The significant costs of a reactor transient are the repair costs resulting from severe damage to plant equipment, the cost of scrams (the actions the system is designed to take to avoid safety risks), US NRC fines, negative publicity, utility rates and revenues. It is estimated that the Three Mile Island-2 accident cost the US over $100 billion in nuclear plant delays and cancellations, more expensive fuel, oil imports, backfits, bureaucratic, administrative and legal costs, and lost productivity

  7. Safety and environmental aspects of deuterium--tritium fusion power plants: work shop summary

    International Nuclear Information System (INIS)

    1978-05-01

    In September of 1977 a workshop was held on the safety and environmental aspects of fusion power plants to consider potential safety and environmental problems of fusion power plants and to reveal solutions or methods of solving those problems. The objective was to promote incorporation of safety and environmental protection into reactor design, thereby reducing the expense and delay of backfitting safety systems after reactor designs are complete. A dialogue was established between fusion reactor designers and safety and environmental researchers. Four topics, each with several subdivisions, were selected for discussion: radiation exposure, accidents, environmental effects, and plant safety. For each topic, discussion focused on the significance of the problem, and adequacy of current technology to solve the problem, design solutions available and research needed to solve the problem

  8. Nuclear power plant licensing and supervision in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Gehrhardt, H.J.; Gottschalk, P.A.

    1991-01-01

    This paper briefly describes nuclear power plant licensing and supervision in the Federal Republic of Germany (FRG). Peculiarities due to the federal structure of the FRG are outlined paying due regard to the long tradition of using consultation by qualified and independent technical experts. The participating authorities, commissions, expert organizations, vendors, utilities and the public as well as their respective competences are mentioned. Also, the hierarchy in nuclear legislation by means of ordinances, administrative regulations, guidelines and technical standards is pointed out. Typical examples are presented. The paper ends in mentioning important items concerning the evaluation of operating experience, recurrent tests, backfitting, lessons learned from the Chernobyl accident, safety research concerning accident management measures, on-site and off-site emergency planning, as well as qualification and occupational training of the responsible shift personnel. (orig.)

  9. LOFT advanced control room operator diagnostic and display system (ODDS)

    International Nuclear Information System (INIS)

    Larsen, D.G.; Robb, T.C.

    1980-01-01

    The Loss-of-Fluid Test (LOFT) Reactor Facility in Idaho includes a highly instrumented nuclear reactor operated by the Department of Energy for the purpose of establishing nuclear safety requirements. The results of the development and installation into LOFT of an Operator Diagnostic and Display System (ODDS) are presented. The ODDS is a computer-based graphics display system centered around a PRIME 550 computer with several RAMTEK color graphic display units located within the control room and available to the reactor operators. Use of computer-based color graphics to aid the reactor operator is discussed. A detailed hardware description of the LOFT data system and the ODDS is presented. Methods and problems of backfitting the ODDS equipment into the LOFT plant are discussed

  10. Pseudostrain representation of multipass excavations in salt

    International Nuclear Information System (INIS)

    Munson, D.E.; Torres, T.M.; Jones, R.L.

    1987-01-01

    Historically, creep closure of excavations in salt have been framed in terms of displacements. However, it appears possible because of the relatively homogeneous nature of salt to introduce a pseudostrain representation, which calls attention to the fact that the displacements are related directly to the strains in the material around the opening. Application of this representation to extensive early-time closure data illustrates the nature of salt response. This paper presents a description of the experimental rooms and explains how early-time mining sequence data were obtained during the multipass excavation. Representative mining sequence displacement data for various size openings are compared and analyzed in terms of pseudostrains. In addition, numerical calculations with a simple back-fitted model of creep are presented. Implications of the results are discussed in the conclusions. 8 refs., 8 figs

  11. Influence of in-plant air pollution control measures on power plant and system operation

    International Nuclear Information System (INIS)

    Kurten, H.

    1990-01-01

    The burning of fossil fuels causes the emission of air pollutants which have harmful environmental impact. Consequently many nations have in the last few years established regulations for air pollution control and have initiated the development and deployment of air pollution control systems in power plants. The paper describes the methods used for reducing particulate, SO 2 and NO x emissions, their application as backfit systems and in new plants, the power plant capacity equipped with such systems in the Federal Republic of Germany and abroad and the additional investment and operating costs incurred. It is to be anticipated that advanced power plant designs will produce lower pollutant emissions and less waste at enhanced efficiency levels. A comparison with power generation in nuclear power plants completes the first part of the paper. This paper covers the impact of the above-mentioned air pollution control measures on unit commitment in daily operation

  12. Environmental information systems - practicable decision aids

    International Nuclear Information System (INIS)

    1988-01-01

    Environmental information systems are classified in documentation systems and environmental planning systems. In environmental information systems emphasis is laid on scientific documentation. Environmental planning systems, on the other hand, involve facts on the state of the environment with respect to the air, noise, water, soil, waste management, the ecology and nature conservation. They can be used as instruments for documenting trends in enviromental pollution and the state of the art in environmental engineering. The relation polluter-environment-enforcement plays a central role for the protection of the environment (integration in terms of the KMSYS). The 'trade and process-specific emissions' system already represents an instrument for the transfer of knowledge in the field of air pollution abatement (see, e.g., Clean Air Technical Code, and the backfitting of existing plants). (DG) [de

  13. Duty and role of Nuclear Regulation Authority facing a crucial moment

    International Nuclear Information System (INIS)

    Asaoka, Mie

    2013-01-01

    Duty of Nuclear Regulation Authority (NRA) was to restore public trust on nuclear regulation spoiled by the Fukushima nuclear accident. How applied such regulation as mandatory back-fitting based on latest knowledge and 40 year operational limit in principle became of great concern. Active faults issue on existing nuclear power station could be a touchstone. Safety side judgment and electric utilities side's proof responsibilities were required as more stringent criteria for active faults. The expert group had been working on assessment of fracture zones under the field survey. Reform of safety regulations should be done based on three important standpoints: (1) not business easiness but public safety was first (2) NRA keeping stance to judge its own safety standard and how NRA ought to be and (3) importance of public disclosure of information and participation in decision-making judging from greatness of public effects caused by nuclear disaster. (T. Tanaka)

  14. Reactor safety: a discussion by officials of the Nuclear Regulatory Commission

    International Nuclear Information System (INIS)

    Anders, W.A.; Rusche, B.C.; Stello, V. Jr.; Minogue, R.B.

    1976-01-01

    William A. Anders, Chairman of the Nuclear Regulatory Commission (NRC), and several senior officials spoke to the Joint Committee on Atomic Energy on the subject of nuclear safety, improvements in reactor plant safety, and quality assurance. The NRC, during its first year of organization, has developed new initiatives to improve safety and safeguards regulations. Anders stressed that NRC is not stifling internal discussion of opposing views, that it has been honest with the public, and that operating reactors are meeting rigorous safety standards. Other speakers discussed comparative safety of old and new reactors. Backfitting of older plants with new features is done when substantial safety protection can be added, but detuning an integrated system is not done indiscriminately. Officials of NRC do not agree with former General Electric employees, who testified that the regulatory procedure is inadequate. Safety improvements since August 28, 1962 and outlines of the review process are included in the Appendixes

  15. Nuclear criticality safety department training implementation

    International Nuclear Information System (INIS)

    Carroll, K.J.; Taylor, R.G.; Worley, C.A.

    1996-01-01

    The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. The NCSD Qualification Program is described in Y/DD-694, Qualification Program, Nuclear Criticality Safety Department This document provides a listing of the roles and responsibilities of NCSD personnel with respect to training and details of the Training Management System (TMS) programs, Mentoring Checklists and Checksheets, as well as other documentation utilized to implement the program. This document supersedes Y/DD-696, Revision 2, dated 3/27/96, Training Implementation, Nuclear Criticality Safety Department. There are no backfit requirements associated with revisions to this document

  16. Seismic safety margins research program. Project I SONGS 1 AFWS Project

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Cummings, G.E.; Wells, J.E.

    1981-01-01

    The seismic qualification requirements of auxiliary feedwater systems (AFWS) of Pressurized Water Reactors (PWR) were developed over a number of years. These are formalized in the publication General Design Criteria (Appendix A to 10CFR50). The full recognition of the system as an engineered safety feature did not occur until publication of the Standard Review Plan (1975). Efforts to determine how to backfit seismic requirements to earlier plants has been undertaken primarily in the Systematic Evaluation Program (SEP) for a limited number of operating reactors. Nuclear Reactor Research (RES) and NRR have requested LLNL to perform a probabilistic study on the AFWS of San Onofre Nuclear Generating Station (SONGS) Unit 1 utilizing the tools developed by the Seismic Safety Margins Research Program (SSMRP). The main objectives of this project are to: identify the weak links of AFWS; compare the failure probabilities of SONGS 1 and Zion 1 AFWS: and compare the seismic responses due to different input spectra and design values

  17. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    1990-11-01

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  18. Implementation of the obligations of the Convention on Nuclear Safety - 6th national report of Switzerland to the Convention in accordance with its article 5

    International Nuclear Information System (INIS)

    2013-08-01

    After a short description of Switzerland as a state in the middle of Europe and of its political organization, the report explains the development of the nuclear power from the first experimental reactor in 1957. Presently five nuclear power plants (NPP) are operating in Switzerland, producing about 40% of the electricity consumption of the country, the rest being produced essentially by hydroelectric plants. As the first regulatory authority, the Swiss Federal Nuclear Safety Commission was set up in 1960, which evolved to the Swiss Nuclear Safety Inspectorate (ENSI). Switzerland signed the Convention on Nuclear Safety (CNS) which came into force at the end of 1996. Since there, Switzerland has prepared and submitted the country reports for the regular Review Meetings of Contracting Countries. This 6th report by ENSI provides an update on compliance with CNS obligations. It gives consideration to issues that aroused particular interest at the 5th meeting and at the extraordinary meeting dedicated to the consequences of the accident at Fukushima Daiichi. Shortly after the accident at Fukushima Daiichi, the Swiss government has decided to phase out nuclear energy; existing plants will continue to operate as long as they are safe. In Switzerland, on-going activities regarding safety assessment of the different stages in the lifetime of nuclear installations consist of periodic assessments and assessments of long-term operation for existing Swiss NPPs. Such assessments have been performed for two Swiss NPPs (Beznau NPP and Muehleberg NPP) which have been in commercial operation for over 40 years. A detailed examination demonstrated that the conditions for the taking out of service of an NPP are not and will not be reached by these two plants within the next 10 years. Nevertheless, it is mandatory to continue with the scheduled ageing management, maintenance and backfitting activities. After the Fukushima accident, additional safety reviews were performed. All Swiss

  19. Big Rock Point: 35 years of electrical generation

    International Nuclear Information System (INIS)

    Petrosky, T.D.

    1998-01-01

    On September 27, 1962, the 75 MWe boiling water reactor, designed and built by General Electric, of the Big Rock Point Nuclear Power Station went critical for the first time. The US Atomic Energy Commission (AEC) and the plant operator, Consumers Power, had designed the plant also as a research reactor. The first studies were devoted to fuel behavior, higher burnup, and materials research. The reactor was also used for medical technology: Co-60 radiation sources were produced for the treatment of more than 120,000 cancer patients. After the accident at the Three Mile Island-2 nuclear generating unit in 1979, Big Rock Point went through an extensive backfitting phase. Personnel from numerous other American nuclear power plants were trained at the simulator of Big Rock Point. The plant was decommissioned permanently on August 29, 1997 after more than 35 years of operation and a cumulated electric power production of 13,291 GWh. A period of five to seven years is estimated for decommissioning and demolition work up to the 'green field' stage. (orig.) [de

  20. CANDU safety management in Pakistan. A status report

    International Nuclear Information System (INIS)

    Mazhar Hasan, S.; Badshah Hussain, S.; Mirza, K.F.; Siddiqui, Z.H.

    1997-01-01

    The overall safety performance of KANUPP against these requirements has been quite good over the past 25 years. But the phenomena of equipment aging, equipment absolescence and evolution of nuclear safety standards, faced by all older NPPs, were aggravated for KANUPP by complete technological isolation from the vendor country for more than 14 years, When it became possible following international attention in 1990, an IAEA sponsored project titled 'Safe Operation of KANUPP (SOK)' was started to assess and ensure compliance to the contemporary internationally acceptable level of safety, leading to a prioritized and Integrated Safety Review Master Plan (ISARMAP) implemented under the supervision of an international Steering Committee. Fortunately, the work done so far has indicated good overall equipment condition, effective obsolescence measures, adequate operational safety practices, and adequate design safety using up-to-date analytical methods. Further detailed analyses and improvements are continuing, to avoid the future potential for an unacceptable level of safety. Difficulties in applying modern safety design standards to backfits are common to older NPPs. 13 refs

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

    International Nuclear Information System (INIS)

    Eckardt, B.

    1991-01-01

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

  2. Reactor risk reference document: Main report: Draft for comment

    International Nuclear Information System (INIS)

    1987-02-01

    The Reactor Risk Reference Document, NUREG-1150, provides the results of major risk analyses for five different US light-water reactors (Surry, Zion, Sequoyah, Peach Bottom, and Grand Gulf) using state-of-the-art methods. The broad base of probabilistic risk information contained in this document is intended to provide a data base and insights to be used in a number of regulatory applications. It is anticipated that these regulatory actions will include implementation of the NRC Severe Accident Policy Statement, implementation of NRC safety goal policy, consideration of the NRC Backfit Rule, evaluation and possible revision of regulations or regulatory requirements for emergency preparedness, plant siting, and equipment qualification, and establishment of risks-oriented priorities for allocating agency resources. This report has been published in draft form. For the plants analyzed, this document describes the major factors related to internally initiated events that contribute to severe core damage, frequencies and related uncertainty ranges of severe core damage events, the major factors and severe accident phenomena that could lead to containment failure, the conditional probabilities and uncertainty ranges of early containment failure, the consequences and risks of severe accidents, including the sensitivity of these risks to factors such as evacuation or sheltering measures, comparisons of the risks with NRC safety goals, and cost and risk-reduction analyses of plant-specific measures that could reduce risk from severe accidents

  3. Surry nuclear power station: 25th anniversary

    International Nuclear Information System (INIS)

    Benson, D.L.

    1998-01-01

    Virginia Power is one of the ten biggest electricity utilities in the United States of America. In 1972, the Surry-1 nuclear generating unit, equipped with an 850 MWe pressurized water reactor from Westinghouse, was accepted into commercial operation. Unit-2 followed in 1973. The North Anna plant is equipped with two 950 MWe PWR commissioned in 1978 and 1980, respectively. The four units together supply roughly one third of the electric power of the grid system in Virginia. They convert nuclear energy into electric power in an economic way: capacity utilization averaged over five years amounted to 90%, and the generating costs were 1.2 cents per kilowatthour. In 1996, the operator began to make use of the experience accumulated in running his plants when backfitting the three generating units on the Millstone site, which are currently out of operation. An agreement on cooperation to this effect was signed by the two utilities, Virginia Power and Northeast Nuclear Energy Company. As a consequence of deregulation of the US electricity market it may be economically preferable to buy electric power instead of generating it in-house. (orig.) [de

  4. Final report, BWR drywell debris transport Phenomena Identification and Ranking Tables (PIRTs)

    International Nuclear Information System (INIS)

    Wilson, G.E.; Boyack, B.E.; Leonard, M.T.; Williams, K.A.; Wolf, L.T.

    1997-09-01

    The Nuclear Regulatory Commission has issued a Regulatory Bulletin and accompanying Regulatory Guide (1.82, Rev. 2) which requires licensees of boiling water reactors to develop a specific plan of action (including hardware backfits, if necessary) to preclude the possibility of early emergency core cooling system strainer blockage following a postulated loss-of-coolant-accident. The postulated mechanism for strainer blockage is destruction of piping insulation in the vicinity of the break and subsequent transport of fragmented insulation to the wetwell. In the absence of more definitive information, the Regulatory Guide recommends that licensees assume a drywell debris transport fraction of 1.0. Accordingly, the Nuclear Regulatory Commission initiated research focused toward developing a technical basis to provide insights useful to regulatory oversight of licensee submittals associated with resolution of the postulated strainer blockage issue. Part of this program was directed towards experimental and analytical research leading to a more realistic specification of the debris transport through the drywell to the wetwell. To help focus this development into a cost effective effort, a panel, with broad based knowledge and experience, was formed to address the relative importance of the various phenomena that can be expected in plant response to postulated accidents that may produce strainer blockage. The resulting phenomena identification and ranking tables reported herein were used to help guide research. The phenomena occurring in boiling water reactors drywells was the specific focus of the panel, although supporting experimental data and calculations of debris transport fractions were considered

  5. Nonparametric modeling and analysis of association between Huntington's disease onset and CAG repeats.

    Science.gov (United States)

    Ma, Yanyuan; Wang, Yuanjia

    2014-04-15

    Huntington's disease (HD) is a neurodegenerative disorder with a dominant genetic mode of inheritance caused by an expansion of CAG repeats on chromosome 4. Typically, a longer sequence of CAG repeat length is associated with increased risk of experiencing earlier onset of HD. Previous studies of the association between HD onset age and CAG length have favored a logistic model, where the CAG repeat length enters the mean and variance components of the logistic model in a complex exponential-linear form. To relax the parametric assumption of the exponential-linear association to the true HD onset distribution, we propose to leave both mean and variance functions of the CAG repeat length unspecified and perform semiparametric estimation in this context through a local kernel and backfitting procedure. Motivated by including family history of HD information available in the family members of participants in the Cooperative Huntington's Observational Research Trial (COHORT), we develop the methodology in the context of mixture data, where some subjects have a positive probability of being risk free. We also allow censoring on the age at onset of disease and accommodate covariates other than the CAG length. We study the theoretical properties of the proposed estimator and derive its asymptotic distribution. Finally, we apply the proposed methods to the COHORT data to estimate the HD onset distribution using a group of study participants and the disease family history information available on their family members. Copyright © 2013 John Wiley & Sons, Ltd.

  6. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  7. Criteria for successful core reflood under severe accident conditions

    International Nuclear Information System (INIS)

    Hering, W.; Homann, Ch.

    2004-01-01

    In present German nuclear power plants, safety is enhanced by prescription of preemptive measures and accident mitigation measures. If preemptive measures (e.g. backfitting, improved safety procedures) should fail, mitigative procedures are foreseen to take credit of all safety-related systems available on site. As usual in daily life, no guarantee of complete success of accident management measures can be given, because the success of a core reflood depends essentially on the actual core state and its history, system pressure, and the injection rate of the activated reflood system. In the present work, available experimental data on core reflood are reviewed to define characteristic regimes and dependencies as well as identifying areas where experimental data are lacking, e.g. reflooding of large in-core debris/pool configurations. A reflood map is proposed based on core state and reflood mass flow rate. Common features of the behavior are deduced on the assumption that non-prototypic facility-based effects can be excluded. (authors)

  8. Probabilistic safety analysis of Novovoronezh-5. The level-1 study overview and findings

    International Nuclear Information System (INIS)

    Lioubarski, A.; Kouzmina, I.; Volkovitski, S.; Samokhine, G.; Berg, T.; Bredova, V.; Joukova, E.; Rozine, V.; Smoutnev, V.

    1997-01-01

    Within the Russian-Swiss Swisrus project, a stage 1 probabilistic safety analysis (PSA) for internally initiated events was carried out for the Novovoronezh-5 nuclear generating unit. The real purpose of the project was the transfer to technical know-how in the field of PSA on the basis of a plant-specific analysis. The study was conducted by scientists of the Scientific and Engineering Center for Nuclear and Radiation Safety (SEC NRS) of the Russian Nuclear Safety Authority, GAN, in close cooperation with experts from the plant. A team headed by the Swiss Central Department for the Safety of Nuclear Installations, HSK, followed the work performed by the Russian scientists, checked, and commented upon, the results, gave instructions and passed on information. When required, workshops were organized on special subjects. The final results and findings were subjected to close scrutiny. The results of the study completed in March 1997 after two and a half years of work have been summarized in a comprehensive final report. The most important conclusions, findings, necessary model improvements, and planned backfitting measures in the plant are presented and discussed. A follow-on project has already been approved and is to be completed by mid-2000. The most important topics to be covered are the application of the PSA model to the plant ('Living PSA'); PSA for external events, including fire and internal flooding; and stage 2 PSA to evaluate containment functioning in major accidents. (orig.) [de

  9. Application of nuclear power station design criteria to non-nuclear installations

    International Nuclear Information System (INIS)

    Regan, J.D.; Hughes, D.J.

    1989-01-01

    The nuclear industry is multi faceted, in that it includes large and complex chemical plants, a large number of different types of nuclear power stations, and on shore ship maintenance facilities, each with its own unique problems. Since the early days the industry has been aware of the additional problem which is superimposed on what may be classed as traditional fire risks, that is, the risk of an uncontrolled release of radioactivity. This has led to the development of sophisticated fire prevention and control techniques which are applied to new plants, and to the backfitting of older plants. The techniques of analysis, design and operation can be applied to both nuclear and non-nuclear installations. Passive protection is preferred backed up by active techniques. Segregation of essential plant to increase the probability of sufficient surviving to ensure safety systems operate and the provision of smoke free, protected escape routes are important aspects of layout and design. Reliability assessments, venting of smoke and hot gases, fire severity analysis, application of mathematical models contribute to the final design to protect against fires. Experiences built up in the fire fighting profession is integrated into the numerical approach by frequent involvement of the local Fire Officers at each stage of the design and layout of installations. (author)

  10. MOX manufacturing perspectives in a fast growing future and the MELOX plant

    International Nuclear Information System (INIS)

    Bekiarian, A.; Le Bastard, G.

    1991-01-01

    The potential MOX fuel market will grow regularly in the nineties. In view of satisfying the needs of the market, mixed-oxide fuel manufacturers have a strong incentive to increase the capacity of existing facilities and to build new ones. The Belgonucleaire plant at Dessel has been in operation since 1973. It has been backfitted up to a capacity of 35 t/y of LWR fuel which is now fully available. To satisfy the need of MOX fuel it was equally decided to adapt facilities in Cadarache where a production line, with a capacity of 15 t/y, is now delivering its production. But planned production up to the end of the century implies further increases in manufacturing capacities : MELOX, a plant for 120 t/y is under construction on the COGEMA site of Marcoule as well as a further expansion of Belgonucleaire plant at Dessel (P1) is studied to reach 70 t/y on this site. Similar developments are also planned by SIEMENS for a new manufacturing capability at Hanau (Germany). MELOX as well as all the new facilities have to get high levels of safety concerning environment and personnel. This leads to largely automated operations, and a particular care for waste treatment. (author)

  11. Current research and development activities on fission products and hydrogen risk after the accident at Fukushima Daiichi Nuclear Power Station

    Directory of Open Access Journals (Sweden)

    Takeshi Nishimura

    2015-02-01

    Full Text Available After the Fukushima Daiichi nuclear power plant (NPP accident, new regulatory requirements were enforced in July 2013 and a backfit was required for all existing nuclear power plants. It is required to take measures to prevent severe accidents and mitigate their radiological consequences. The Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R has been conducting numerical studies and experimental studies on relevant severe accident phenomena and countermeasures. This article highlights fission product (FP release and hydrogen risk as two major areas. Relevant activities in the S/NRA/R are briefly introduced, as follows: 1. For FP release: Identifying the source terms and leak mechanisms is a key issue from the viewpoint of understanding the progression of accident phenomena and planning effective countermeasures that take into account vulnerabilities of containment under severe accident conditions. To resolve these issues, the activities focus on wet well venting, pool scrubbing, iodine chemistry (in-vessel and ex-vessel, containment failure mode, and treatment of radioactive liquid effluent. 2. For hydrogen risk: because of three incidents of hydrogen explosion in reactor buildings, a comprehensive reinforcement of the hydrogen risk management has been a high priority topic. Therefore, the activities in evaluation methods focus on hydrogen generation, hydrogen distribution, and hydrogen combustion.

  12. Use of non standard methods for seismic evaluation of piping systems in existing plants

    International Nuclear Information System (INIS)

    Geraets, L.H.; Lafaille, J.P.; Mignot, P.

    1989-01-01

    Seismic design and analysis of nuclear plant structures, systems and components have requested huge efforts and tremendous costs in the past fifteen years. The extended use of sophisticated, linear response-type methods are responsible for the significant stiffening of the piping systems and the multiplication of supports, and in particular of snubbers. The remedy used against the seismic risk seems to be worse than the disease itself, and safety might be impaired rather than improved. Indeed, stiffening of the system increases the average load level in normal operation, supports do not behave ideally as assumed and snubbers happen to be remarkably unreliable. On the other hand, experience with actual earthquakes shows that industrial facilities designed using very simplistic seismic techniques or even no seismic requirement at all suffer essentially no damage, even in the case of a large earthquake. This paradox challenges the traditional design techniques, and appeals for revised approaches for seismic qualification of piping systems. When the assumption of the occurrence of an earthquake event is made in a plant in operation, which has not been designed using seismic criteria, the use of the standard seismic qualification techniques would still be more questionable. The use of simplified quasi-static techniques, backed when requested with sophisticated time history non linear analyses, allows to assess the seismic adequacy of the safety-related critical piping systems with a minimum backfitting actions

  13. Regulatory risks associated with nuclear safety legislation after Fukushima Daiichi Nuclear Accident in Japan. Focus on legal structure of the nuclear reactor regulation act

    International Nuclear Information System (INIS)

    Tanabe, Tomoyuki; Maruyama, Masahiro

    2016-01-01

    Nuclear safety regulations enforced after Fukushima Daiichi Nuclear Accident under the Nuclear Reactor Regulation Act face the following regulatory problems that involve potential risk factors for nuclear businesses; 1) 'entity based regulation' unable to cope with business cessation or bankruptcy of the entity subject of regulation, 2) potential risk of the Nuclear Regulation Authority's inappropriate involvement in nuclear industry policy beyond their duty, and 3) compliance of backfits under vague regulations. In order to alleviate them, this report, through analyzing these regulatory problems from the view point of sound development of the nuclear industry, proposes the following regulatory reforms; (1) To clarify the rule for industry policy in nuclear regulations and enable the authority, Ministry of Economy, Trade and Industry, to choose most appropriate industrial policy measure. (2) Through establishing safety goals as measures to promote continuous improvement of nuclear safety regulations, to stimulate timely adjustments of the regulations, and to introduce a legal mechanism into the nuclear regulation systems under which validity of administrative law and its application can be checked. (author)

  14. Zebra Mussel Monitoring and Control Guide

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-01

    The Electric Power Research Institute (EPRI) Zebra Mussel Monitoring and Control Guide is a comprehensive compilation of US and European practices as reported in the open literature as of the end of 1992. EPRI considers the guide to be a living' document and will update it periodically in order to provide results of current research on chemical and nonchemical control technologies and utility experiences. The zebra mussel has infested all of the Great Lakes and other major rivers and waterways and is positioned to spread even more to the adjoining river basins. The impact of the zebra mussel on industrial power plantsis as a biofouler that clogs water systems and heat exchangers. This EPRI guideline identifies the zebra mussel, discusses its distribution in the United States, presents the potential threats to power plants, and presents the methods to initiate monitoring and control programs. Both preventive and corrective measures are presented. Preventive measures include various monitoring methods to initiate control techniques. The control techniques include both chemical and nonchemical together with combining techniques. Corrective methods include operational considerations, chemical cleaning, and mechanical/physical cleaning. It also may be possible to incorporate design changes, such as open to closed-loop backfit, backflushing, or pretreatment for closed systems. Various appendices are included that contain specifications to aid utilities in implementing several of the monitoring and control technologies, results of chemical evaluations at Cleveland Electric Illuminating Company plants, and data on the fate of various commercial molluscicides.

  15. Cost-benefit considerations in regulatory analysis

    International Nuclear Information System (INIS)

    Mubayi, V.; Sailor, V.; Anandalingam, G.

    1995-10-01

    Justification for safety enhancements at nuclear facilities, e.g., a compulsory backfit to nuclear power plants, requires a value-impact analysis of the increase in overall public protection versus the cost of implementation. It has been customary to assess the benefits in terms of radiation dose to the public averted by the introduction of the safety enhancement. Comparison of such benefits with the costs of the enhancement then requires an estimate of the monetary value of averted dose (dollars/person rem). This report reviews available information on a variety of factors that affect this valuation and assesses the continuing validity of the figure of $1000/person-rem averted, which has been widely used as a guideline in performing value-impact analyses. Factors that bear on this valuation include the health risks of radiation doses, especially the higher risk estimates of the BEIR V committee, recent calculations of doses and offsite costs by consequence codes for hypothesized severe accidents at U.S. nuclear power plants under the NUREG-1150 program, and recent information on the economic consequences of the Chernobyl accident in the Soviet Union and estimates of risk avoidance based on the willingness-to-pay criterion. The report analyzes these factors and presents results on the dollars/person-rem ratio arising from different assumptions on the values of these factors

  16. Evaluation of severe accident risks and the potential for risk reduction: Grand Gulf, Unit 1. Draft for comment, February 1987

    International Nuclear Information System (INIS)

    Amos, C.N.; Benjamin, A.S.; Kunsman, D.M.; Williams, D.C.; Boyd, G.J.; Lewis, S.R.; Smith, L.N.

    1987-04-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a boiling water reactor with a Mark III containment (Grand Gulf, Unit 1). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally low relative to previous studies; for example, most of the uncertainty range is lower than the point estimate of risk for the Peach Bottom plant in the Reactor Safety Study (RSS). However, certain unresolved issues cause the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. These issues include the diesel generator failure rate, iodine and cesium revolatilization after vessel breach and the possibility of reactor vessel pedestal failure caused by core debris attack. Some of the postulated safety options appear to be potentially cost effective for the Grand Gulf power plant, particularly when onsite accidents costs are included in the evaluation of benefits. Principally these include procedural modifications and relatively inexpensive hardware additions to insure core cooling in the event of a station blackout. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150. (author)

  17. Regulatory analysis for the resolution of generic issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment

    International Nuclear Information System (INIS)

    Woods, H.W.

    1993-10-01

    Actuation of Fire Protection Systems (FPS) in Nuclear Power Plants have resulted in adverse interactions with equipment important to safety. Precursor operational experience has shown that 37% of all FPS actuations damaged some equipment, and 20% of all FPS actuations have resulted in a plant transient and reactor trip. On an average 0.17 FPS actuations per reactor year have been experienced in nuclear power plants in this country. This report presents the regulatory analysis for GI-57, ''Effects of Fire Protection System Actuation on Safety-Related Equipment''. The risk reduction estimates, cost/benefit analyses, and other insights gained during this effort have shown that implementation of the recommendations contained in this report can significantly reduce risk, and that these improvements can be warranted in accordance with the backfit rule, 10 CFR 50.109(a)(3). However, plant specific analyses are required in order to identify such improvements. Generic analyses can not serve to identify improvements that could be warranted for individual, specific plants. Plant specific analyses of the type needed for this purpose are underway as part of the Individual Plant Examination of External Events (IPEEE) program

  18. The CEGB/SSEB response to Recommendation 17 in the Environment Committee's Report on Radioactive Waste. V.1

    International Nuclear Information System (INIS)

    1986-11-01

    The first report from the Environment Committee concerning radioactive waste was published on 12th March 1986. Recommendation 17 of the Committee's report asked the CEGB and SSEB (the Home Boards) to carry out and publish an analysis of the costs of backfitting dry stores to Magnox stations and compare this with the costs of reprocessing, vitrification and subsequent storage of vitrified HLW. In addition the Committee asked that the Home Boards should examine the feasibility of drying Magnox spent fuel once it had been wet in the cooling ponds. This report represents the Home Boards' response to Recommendation 17. In addition, in order to provide a comprehensive economic comparison, consideration has also been given to the likely range of costs for treatment and final disposal of Magnox spent fuel. In carrying out this study the Home Boards have assessed the technical feasibility, costs and likely timescales associated with establishing new all-dry discharge routes on each of the individual Magnox stations and constructing dry storage facilities suitable for storing Magnox fuel for up to 100 years. (author)

  19. The materials concept in German light water reactors. A contribution to plant safety, economic performance and damage prevention

    International Nuclear Information System (INIS)

    Ilg, Ulf

    2008-01-01

    Major decisions taken as early as in the planning and construction phases of nuclear power plants may influence overall plant life. Component quality at the beginning of plant life is determined very much also by a balanced inclusion of the 'design, choice of materials, manufacturing and inspection' elements. One example of the holistic treatment of design, choice of material, and manufacture of important safety-related components in pressurized water reactors is the reactor pressure vessel (RPV) in which the ferritic compound tubes, with inside claddings, for the control rod drive nozzles are screwed into the vessel top. Also the choice of Incoloy 800 for the steam generator tubes, and the design of the main coolant pipes with inside claddings as seamless pipe bends / straight pipes with integrated nozzles connected to mixed welds with austenitic pipes are other special design features of the Siemens/KWU plants. A demonstrably high quality standard by international comparison to this day has been exhibited by the austenitic RPV internals of boiling water reactors, which were made of a low-carbon Nb-stabilized austenitic steel grade by optimum manufacturing technologies. The same material is used for backfitting austenitic pipes. Reliable and safe operation of German nuclear power plants has been demonstrated for more than 4 decades. One major element in this performance is the materials concept adopted in Germany also in the interest of damage prevention. (orig.)

  20. Probabilistic safety assessment past, present and future. An IAEA perspective

    International Nuclear Information System (INIS)

    Lederman, L.; Niehaus, F.; Tomic, B.

    1996-01-01

    Despite the high level of development that probabilistic safety assessment (PSA) methods have reached, a number of issues place constraints on its use in supporting decision making on safety matters. A recent publication of the International Nuclear Safety Advisory Group (INSAG) represents an important step in reaching international consensus on the use of PSA. PSA is ''strongly encouraged'' by INSAG; however, it is noted that ''PSA methodology is not sufficiently mature for its present status to be frozen''. The main aspects of the report are discussed in this paper. The paper next discusses three main categories of PSA application, namely the adequacy of design and procedures, optimization of operational activities and regulatory applications. For each of the applications, the objectives, specific modelling requirements and the prospects for implementation are presented. Consistent with its statutory functions, an important aspect of the work of the IAEA is to reach international consensus on the possibilities of and limitations on the use of PSA methods. Whereas past efforts have been concentrated on promotion and assistance to perform Level 1 PSAs, work is now extending with emphasis on PSA applications, Level 2 and Level 3 analysis, external events and shutdown risks. The main elements of IAEA's PSA Programme are discussed. Finally some challenges related to the use of PSA in the backfitting of nuclear power plants in Eastern Europe and countries of the former USSR are addressed. (orig.)

  1. The development of engineered safeguards for nuclear power plants in the political and technical environment in the Federal Republic of Germany since 1955; Die Entwicklung der Sicherheitstechnik fuer Kernkraftwerke im politischen und technischen Umfeld der Bundesrepublik Deutschland seit dem Jahr 1955

    Energy Technology Data Exchange (ETDEWEB)

    Laufs, P. [Stuttgart Univ. (Germany). Philosophische-Historische Fakultaet

    2007-01-15

    The safety of nuclear power plants is determined largely by the integrity of the internally pressurized coolant containment system. The highly radioactive materials (fission products) generated within this pressurized containment (primary system) in the reactor core during nuclear power plant operation constitute an extremely great potential hazard. Catastrophic failure of the primary system, and the release into the environment of the radioactive inventory, must be avoided at all costs. Because of the high coolant pressure and the high power density, pressurized water reactors (PWR) impose particularly strict requirements with respect to reactor safety. German nuclear power plants equipped with light water reactors enjoy the reputation of being among the safest plants in the world. This frequent statement is justified in the light of the research and development work performed jointly by industry, government agencies, science, and expert bodies between the 1960s and the 1990s. The research projects, which implied considerable financial expenditures, their internationally acknowledged results, and the resultant additional backfitting measures conducted in German nuclear power plants at many billions of expenditures, were hardly noticed by the German body politic. (orig.)

  2. Uranium resource utilization improvements in the once-through PWR fuel cycle

    International Nuclear Information System (INIS)

    Matzie, R.A.

    1980-04-01

    In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U 3 O 8 consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout

  3. Proposal strategy and policy on nuclear safety for no-more severe accidents

    International Nuclear Information System (INIS)

    2013-01-01

    Following the outspoken advice saying 'scientists and engineers concerning with nuclear power promotion and safety should be responsible for clarifying how preventable or what measures should be needed to prevent severe accidents occurring at Fukushima Daiichi nuclear power plants (NPPs)', committee on prevention of severe accidents at NPPs was established by relevant nuclear scientists and engineers involved so as to discuss basic issues to be solved from scientific and technical viewpoints. Based on the review of 'defense in depth' concept and accident analysis at Fukushima nuclear accident, four major proposals and six supplements to be established were identified such as: (1) finding mechanism of beyond imagination events for natural disaster, terrorism, and internal events, (2) reform of comprehensive safety standards and guidelines with performance basis easy to reflect latest knowledge and technology as 'back-fitting', (3) severe accidents measures, their validation, and drilling on accident management to advance procedures and develop human resources, and (4) risk communications and public disclosure of information. This article described backgrounds of committee's proposals on nuclear safety for no-more severe accidents. (T. Tanaka)

  4. Peer evaluation and some valuable lessons

    International Nuclear Information System (INIS)

    Holt, A.G.

    1991-01-01

    In the mid 1980s there were some signs that Ontario Hydro's nuclear program performance was deteriorating. Such signs included increased maintenance backlog, increased number of jumpers, decreased capacity factors and increasing regulatory concerns. Factors influencing this deterioration were: (a) Pressure tube creep and hydriding rates were excessive leading to increased reactor maintenance and early pressure tube replacement in Pickering NGS-A and Bruce NGS-A. (b) Preventive maintenance was reduced to a minimum owing to manpower and budget restraints. This led to more forced outages, deratings and breakdown maintenance as the urgent was dealt with rather than the important. (c) New systems were installed in the older units, Pickering NGS-A and Bruce NGS-A, in order to backfit safety related system improvements principally to meet increased regulatory requirements. This put additional strain on tight resources to assist with the installation, commissioning, testing and maintenance of these systems that generally increased the complexity of units. Again this led to a reduction of preventive maintenance

  5. Interdisciplinary study of the influence on effectiveness of catalytic hydrogen recombiners of operating conditions in the reactor containment

    International Nuclear Information System (INIS)

    Kelm, S.; Reinecke, E.A.; Schoppe, L.; Dornseiffer, J.; Leistner, F.; Juehe, S.

    2008-01-01

    At the Emsland nuclear power station, a total of 58 autocatalytic hydrogen recombiners were backfitted in 1999 as an additional measure of risk reduction in connection with major hydrogen releases after events going beyond the design basis. Annual in-service inspections after 2002 revealed that some of the catalyst sheets developed startup delays and marked evolutions of smoke and smell. Recombiners not meeting the inspection criterion were completely regenerated as a measure of precaution. A preventive study was conducted jointly with institutes of the Juelich Research Center and the Aachen Technical University to analyze the composition of the deposits, which was then compared with the chemical characteristics of potential sources in the reactor containment. At the same time, the influence on effectiveness of the catalyst sheets was examined. On the basis of a random evaluation of the in-service inspection logs of the past few years, representative samples were taken whose startup behavior and operating characteristics were studied in a test rig alongside chemical analyses so as to allow a correlation to be established between the analytical findings and the catalytic activity of the samples. The findings made allowed internal sources of the catalyst deposits to be excluded. The impurities are introduced with the outside air. As a consequence, the air ducts in the vicinity of the respective recombiners were inspected and optimization steps were taken in connection with in-service inspections and regeneration procedures. (orig.)

  6. German nuclear codes revised: comparison with approaches used in other countries

    International Nuclear Information System (INIS)

    Raetzke, C.; Micklinghoff, M.

    2005-01-01

    The article deals with the plan of the German Federal Ministry for the Environment (BMU) to revise the German set of nuclear codes, and draws a comparison with approaches pursued in other countries in formulating and implementing new requirements imposed upon existing plants. A striking feature of the BMU project is the intention to have the codes reflect the state of the art in an entirely abstract way irrespective of existing plants. This implies new requirements imposed on plant design, among other things. However, the state authorities, which establish the licensing conditions for individual plants in concrete terms, will not be able to apply these new codes for legal reasons (protection of vested rights) to the extent in which they incorporate changes in safety philosophy. Also the procedure adopted has raised considerable concern. The processing time of two years is inordinately short, and participation of the public and of industry does not go beyond the strictly formal framework of general public participation. In the light of this absence of quality assurance, it would be surprising if this new set of codes did not suffer from considerable deficits in its contents. Other countries show that the BMU is embarking on an isolated approach in every respect. Elsewhere, backfitting requirements are developed carefully and over long periods of time; they are discussed in detail with the operators; costs and benefits are weighted, and the consequences are evaluated. These elements are in common to procedures in all countries, irrespective of very different steps in detail. (orig.)

  7. Regulatory analysis technical evaluation handbook. Final report

    International Nuclear Information System (INIS)

    1997-01-01

    The purpose of this Handbook is to provide guidance to the regulatory analyst to promote preparation of quality regulatory analysis documents and to implement the policies of the Regulatory Analysis Guidelines of the US Nuclear Regulatory Commission (NUREG/BR-0058 Rev. 2). This Handbook expands upon policy concepts included in the NRC Guidelines and translates the six steps in preparing regulatory analyses into implementable methodologies for the analyst. It provides standardized methods of preparation and presentation of regulatory analyses, with the inclusion of input that will satisfy all backfit requirements and requirements of NRC's Committee to Review Generic Requirements. Information on the objectives of the safety goal evaluation process and potential data sources for preparing a safety goal evaluation is also included. Consistent application of the methods provided here will result in more directly comparable analyses, thus aiding decision-makers in evaluating and comparing various regulatory actions. The handbook is being issued in loose-leaf format to facilitate revisions. NRC intends to periodically revise the handbook as new and improved guidance, data, and methods become available

  8. Workshop on environmental qualification of electric equipment

    Energy Technology Data Exchange (ETDEWEB)

    Lofaro, R.; Gunther, W.; Villaran, M.; Lee, B.S.; Taylor, J. [comps.] [Brookhaven National Lab., Upton, NY (United States)

    1994-05-01

    Questions concerning the Environmental Qualification (EQ) of electrical equipment used in commercial nuclear power plants have recently become the subject of significant interest to the US Nuclear Regulatory Commission (NRC). Initial questions centered on whether compliance with the EQ requirements for older plants were adequate to support plant operation beyond 40 years. After subsequent investigation, the NRC Staff concluded that questions related to the differences in EQ requirements between older and newer plants constitute a potential generic issue which should be evaluated for backfit, independent of license renewal activities. EQ testing of electric cables was performed by Sandia National Laboratories (SNL) under contract to the NRC in support of license renewal activities. Results showed that some of the environmentally qualified cables either failed or exhibited marginal insulation resistance after a simulated plant life of 20 years during accident simulation. This indicated that the EQ process for some electric cables may be non-conservative. These results raised questions regarding the EQ process including the bases for conclusions about the qualified life of components based upon artificial aging prior to testing.

  9. Workshop on environmental qualification of electric equipment

    International Nuclear Information System (INIS)

    Lofaro, R.; Gunther, W.; Villaran, M.; Lee, B.S.; Taylor, J.

    1994-05-01

    Questions concerning the Environmental Qualification (EQ) of electrical equipment used in commercial nuclear power plants have recently become the subject of significant interest to the US Nuclear Regulatory Commission (NRC). Initial questions centered on whether compliance with the EQ requirements for older plants were adequate to support plant operation beyond 40 years. After subsequent investigation, the NRC Staff concluded that questions related to the differences in EQ requirements between older and newer plants constitute a potential generic issue which should be evaluated for backfit, independent of license renewal activities. EQ testing of electric cables was performed by Sandia National Laboratories (SNL) under contract to the NRC in support of license renewal activities. Results showed that some of the environmentally qualified cables either failed or exhibited marginal insulation resistance after a simulated plant life of 20 years during accident simulation. This indicated that the EQ process for some electric cables may be non-conservative. These results raised questions regarding the EQ process including the bases for conclusions about the qualified life of components based upon artificial aging prior to testing

  10. Application of advanced technology to LMR control

    International Nuclear Information System (INIS)

    Lindsay, R.W.

    1989-01-01

    Key issues must be resolved to preserve the nuclear option; including new considerations for safety, economics, waste, transportation, diversion, etc. The programs at the Experimental Breeder Reactor II (EBR-II) are now carefully focused to provide answers to the above concerns in connection with the Integral Fast Reactor program at Argonne. Safety features that are inherent in plant design, coupled with automating plant control to help achieve the above objectives are more than just an issue of installing controllers and exotic algorithms, they include the complete integration of plant design, control strategy, and information presentation. Current technology development, both at Argonne and elsewhere includes efforts relating to the use of Artificial Intelligence, sensor/signal validation in many forms, pattern recognition, optimal develop and/or adopt promising technologies, and integrate them into an operating power plant for proof of value. After they have proven useful at EBR-II, it is expected that they can be incorporated into advanced designs such as PRISM and/or included in backfit activities as well. 6 refs

  11. Containment hydrogen and atmosphere activity control to mitigate severe accidents in VVERs and Western PWRs. Design and status of implementation

    International Nuclear Information System (INIS)

    Feuerbach, R.

    2002-01-01

    For accident management nuclear power plants in Europe have been or will be back-fitted with supplementary systems for monitoring the containment hydrogen concentration, for the early removal and reduction of hydrogen and filtered venting systems to retain radioactive aerosols and iodine. The hydrogen monitoring system (HMS) provides the information of local H 2 concentration in the containment during DBA and severe accident situations. The new HMS contains of overall H 2 -sensors and is installed inside the confinement. It provides continuos information about the local and temporal distribution of hydrogen, reported directly to the Emergency Response Team in case of severe accident. The hydrogen Reduction System (HRS) consists of several Passive Autocatalytic Recombiners (PAR) located in several compartments in the containment. The number of PARs to be installed depends on the type of NPP, structure of containment and the investigated accident scenario e.g. DBA conditions - approx. 6 to 20 PARs; severe accident conditions - 20-60 PARs). In case of severe accident it does not need any operator actions. The Filtered Venting System (FVS) is is especially important for WWER-440/230 maintaining sub atmospheric pressure in the confinement. For severe accident the on-site Emergency Response Team has to take the necessary strategic decisions for containment depressurization via the FVS

  12. Cost-benefit considerations in regulatory analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mubayi, V.; Sailor, V.; Anandalingam, G.

    1995-10-01

    Justification for safety enhancements at nuclear facilities, e.g., a compulsory backfit to nuclear power plants, requires a value-impact analysis of the increase in overall public protection versus the cost of implementation. It has been customary to assess the benefits in terms of radiation dose to the public averted by the introduction of the safety enhancement. Comparison of such benefits with the costs of the enhancement then requires an estimate of the monetary value of averted dose (dollars/person rem). This report reviews available information on a variety of factors that affect this valuation and assesses the continuing validity of the figure of $1000/person-rem averted, which has been widely used as a guideline in performing value-impact analyses. Factors that bear on this valuation include the health risks of radiation doses, especially the higher risk estimates of the BEIR V committee, recent calculations of doses and offsite costs by consequence codes for hypothesized severe accidents at U.S. nuclear power plants under the NUREG-1150 program, and recent information on the economic consequences of the Chernobyl accident in the Soviet Union and estimates of risk avoidance based on the willingness-to-pay criterion. The report analyzes these factors and presents results on the dollars/person-rem ratio arising from different assumptions on the values of these factors.

  13. Seismic and tsunami safety margin assessment

    International Nuclear Information System (INIS)

    2013-01-01

    Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

  14. Tsunami hazard

    International Nuclear Information System (INIS)

    2013-01-01

    Tohoku Earthquake Tsunami on 11 March, 2011 has led the Fukushima Daiichi nuclear power plant to a serious accident, which highlighted a variety of technical issues such as a very low design tsunami height and insufficient preparations in case a tsunami exceeding the design tsunami height. Lessons such as to take measures to be able to maintain the important safety features of the facility for tsunamis exceeding design height and to implement risk management utilizing Probabilistic Safety Assessment are shown. In order to implement the safety assessment on nuclear power plants across Japan accordingly to the back-fit rule, Nuclear Regulatory Commission will promulgate/execute the New Safety Design Criteria in July 2013. JNES has positioned the 'enhancement of probabilistic tsunami hazard assessment' as highest priority issue and implemented in order to support technically the Nuclear Regulatory Authority in formulating the new Safety Design Criteria. Findings of the research had reflected in the 'Technical Review Guidelines for Assessing Design Tsunami Height based on tsunami hazards'. (author)

  15. Generalized partial linear varying multi-index coefficient model for gene-environment interactions.

    Science.gov (United States)

    Liu, Xu; Gao, Bin; Cui, Yuehua

    2017-03-01

    Epidemiological studies have suggested the joint effect of simultaneous exposures to multiple environments on disease risk. However, how environmental mixtures as a whole jointly modify genetic effect on disease risk is still largely unknown. Given the importance of gene-environment (G×E) interactions on many complex diseases, rigorously assessing the interaction effect between genes and environmental mixtures as a whole could shed novel insights into the etiology of complex diseases. For this purpose, we propose a generalized partial linear varying multi-index coefficient model (GPLVMICM) to capture the genetic effect on disease risk modulated by multiple environments as a whole. GPLVMICM is semiparametric in nature which allows different index loading parameters in different index functions. We estimate the parametric parameters by a profile procedure, and the nonparametric index functions by a B-spline backfitted kernel method. Under some regularity conditions, the proposed parametric and nonparametric estimators are shown to be consistent and asymptotically normal. We propose a generalized likelihood ratio (GLR) test to rigorously assess the linearity of the interaction effect between multiple environments and a gene, while apply a parametric likelihood test to detect linear G×E interaction effect. The finite sample performance of the proposed method is examined through simulation studies and is further illustrated through a real data analysis.

  16. Topical issues in nuclear, radiation and radioactive waste safety. Proceedings of an international conference

    International Nuclear Information System (INIS)

    1999-01-01

    The objective of the conference was to foster the exchange of information on topical issues in nuclear, radiation and radioactive waste safety, with the aim of consolidating an international consensus on the current status of these issues, priorities for future work and the need for strengthening international co-operation, including recommendations for the IAEA's future activities. The topical issues were grouped under the following six major headings: safety management; occupational radiation protection - trends and developments; backfitting, upgrading and modernization of nuclear power plants; situations of chronic exposure to residual radioactive materials - decommissioning and rehabilitation and reclamation of land; radiation safety in the distant future - the issue of long tern waste disposal; regulatory strategies. This volume contains the topical issue papers, the keynote presentations, the current issue presentations, the conclusions of the six technical sessions, and the conference chairperson's summary of findings and conclusions. Each of these papers has been provided with an abstract and indexed separately. Individual contributions to this conference have been published separately in the IAEA-TECDOC-1031. A CD-ROM containing contributed papers is attached to this book

  17. Detrimental effects of subsequent increases in the safety and quality requirements using the Grohnde nuclear power station as an example

    International Nuclear Information System (INIS)

    Boettcher, D.

    1983-01-01

    The excellent operational availability and freedom from faults of German nuclear powerstations should give one the courage to take further sensible steps. From the operator's view these include: - Refusal to accept backfitting to a different state of science and technology. Instead of this, orderly introduction of new solutions after careful testing, unless meeting an emergency requires immediate action. - Further support of efforts at standardization of the industry with the possibility of transferring experience. - Reducing multiple inspections (the previous occurrence of multiple inspections in manufacture and erection in a system hides the danger of routine and creeping delegation of responsibility and attention among those concerned). - Limiting the extent of structural and repeat tests to the essential minimum, particularly where there are hold-ups caused during manufacture and erection, which prevent optimum economic construction. - Dispensing with complete documentation of every activity by the applicant, manufacturer, authority and expert. This may contribute to providing proofs for legal processes, but does not contribute to obtaining greater safety. (orig./RW) [de

  18. NUMARC view of license renewal criteria

    International Nuclear Information System (INIS)

    Edwards, D.W.

    1989-01-01

    The Atomic Energy Act and the implementing regulations of the US Nuclear Regulatory Commission (NRC) permit the renewal of nuclear plant operating licenses upon expiration of their 40-year license term. However, the regulatory process by which license renewal may be accomplished and the requirements for the scope and content of renewal applications are yet to be established. On August 29, 1988, the NRC published an Advanced Notice of Proposed Rulemaking regarding the subject of license renewal. This Advanced Notice and the NUREG which it references, NUREG-1317, Regulatory Options for Nuclear Plant License Renewal, provide the most recent regulatory thought on this issue. The basic issue addressed by NUREG-1317 is the definition of an adequate licensing basis for the renewal of a plant license. The report contemplates three alternatives in this regard. This paper discusses each of these three proposals. The NUMARC NUPLEX Working Group endorses a license renewal process based on a plant's current licensing basis along with an evaluation of the pertinent components, systems, and structures affected by age-related degradation. The NUMARC NUPLEX Working group believes that an appropriate scope for NRC review of the license renewal application should focus on those safety-significant structures systems, and components subject to significant age-related degradation that are not subject to existing recognized effective replacement, refurbishment, or inspection programs. The paper also briefly discusses NUMARC's view of the role of the Backfit Rule in the license renewal process

  19. A human reliability assessment screening method for the NRU upgrade project

    International Nuclear Information System (INIS)

    Bremner, F.M.; Alsop, C.J.

    1997-01-01

    The National Research Universal (NRU) reactor is a 130MW, low pressure, heavy water cooled and moderated research reactor. The reactor is used for research, both in support of Canada's CANDU development program, and for a wide variety of other research applications. In addition, NRU plays an important part in the production of medical isotopes, e.g., generating 80% of worldwide supplies of Molybdenum-99. NRU is owned and operated by Atomic Energy of Canada Ltd. (AECL), and is currently undergoing upgrading as part of AECL's continuing commitment to operate their facilities in a safe manner. As part of these upgrades both deterministic and probabilistic safety assessments are being carried out. It was recognized that the assignment of Human Error Probabilities (HEPs) is an important part of the Probabilistic Safety Assessment (PSA) studies, particularly for a facility whose design predates modern ergonomic practices, and which will undergo a series of backfitted modifications whilst continuing to operate. A simple Human Reliability Assessment (HRA) screening method, looking at both pre- and post-accident errors, was used in the initial safety studies. However, following review of this method within AECL and externally by the regulator, it was judged that benefits could be gained for future error reduction by including additional features, as later described in this document. The HRA development project consisted of several stages; needs analysis, literature review, development of method (including testing and evaluation), and implementation. This paper discusses each of these stages in further detail. (author)

  20. Regulatory analysis technical evaluation handbook. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    The purpose of this Handbook is to provide guidance to the regulatory analyst to promote preparation of quality regulatory analysis documents and to implement the policies of the Regulatory Analysis Guidelines of the US Nuclear Regulatory Commission (NUREG/BR-0058 Rev. 2). This Handbook expands upon policy concepts included in the NRC Guidelines and translates the six steps in preparing regulatory analyses into implementable methodologies for the analyst. It provides standardized methods of preparation and presentation of regulatory analyses, with the inclusion of input that will satisfy all backfit requirements and requirements of NRC`s Committee to Review Generic Requirements. Information on the objectives of the safety goal evaluation process and potential data sources for preparing a safety goal evaluation is also included. Consistent application of the methods provided here will result in more directly comparable analyses, thus aiding decision-makers in evaluating and comparing various regulatory actions. The handbook is being issued in loose-leaf format to facilitate revisions. NRC intends to periodically revise the handbook as new and improved guidance, data, and methods become available.

  1. Potential of low-temperature nuclear heat applications

    International Nuclear Information System (INIS)

    1986-12-01

    At present, more than one third of the fossil fuel currently used is being consumed to produce space heating and to meet industrial needs in many countries of the world. Imported oil still represents a large portion of this fossil fuel and despite its present relatively low price future market evolutions with consequent upward cost revisions cannot be excluded. Thus the displacement of the fossil fuel by cheaper low-temperature heat produced in nuclear power plants is a matter which deserves careful consideration. Technico-economic studies in many countries have shown that the use of nuclear heat is fully competitive with most of fossil-fuelled plants, the higher investment costs being offset by lower production cost. Another point in favour of heat generation by nuclear source is its indisputable advantage in terms of benefits to the environment. The IAEA activity plans for 1985-86 concentrate on information exchange with specific emphasis on the design criteria, operating experience, safety requirements and specifications of heat-only reactors, co-generation plants and existing power plants backfitted for additional heat applications. The information gained up to 1985 was discussed during the Advisory Group Meeting on the Potential of Low-Temperature Nuclear Heat Applications held in the Federal Institute for Reactor Research, Wuerenlingen, Switzerland in September 1985 and, is included in the present Technical Document

  2. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  3. Computerized reactor protection and safety related systems in nuclear power plants. Proceedings of a specialists' meeting. Working material

    International Nuclear Information System (INIS)

    1998-01-01

    Though the majority of existing control and protection systems in nuclear power plants use old analogue technology and design philosophy, the use of computers in safety and safety related systems is becoming a current practice. The Specialists Meeting on ''Computerized Reactor Protection and Safety Related Systems in Nuclear Power Plants'' was organized by IAEA (jointly by the Division of Nuclear Power and the Fuel Cycle and the Division of Nuclear Installation Safety), in co-operation with Paks Nuclear Power Plant in Hungary and was held from 27-29 October 1997 in Budapest, Hungary. The meeting focused on computerized safety systems under refurbishment, software reliability issues, licensing experiences and experiences in implemented computerized safety and safety related systems. Within a meeting programme a technical visit to Paks NPP was organized. The objective of the meeting was to provide an international forum for the presentation and discussion on R and D, in-plant experiences in I and C important to safety, backfits and arguments for and reservations against the digital safety systems. The meeting was attended by 70 participants from 16 countries representing NPPs and utility organizations, design/engineering, research and development, and regulatory organizations. In the course of 4 sessions 25 technical presentations were made. The present volume contains the papers presented by national delegates and the conclusions drawn from the final general discussion

  4. Final report, BWR drywell debris transport Phenomena Identification and Ranking Tables (PIRTs)

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); Boyack, B.E. [Los Alamos National Lab., NM (United States); Leonard, M.T.; Williams, K.A.; Wolf, L.T.

    1997-09-01

    The Nuclear Regulatory Commission has issued a Regulatory Bulletin and accompanying Regulatory Guide (1.82, Rev. 2) which requires licensees of boiling water reactors to develop a specific plan of action (including hardware backfits, if necessary) to preclude the possibility of early emergency core cooling system strainer blockage following a postulated loss-of-coolant-accident. The postulated mechanism for strainer blockage is destruction of piping insulation in the vicinity of the break and subsequent transport of fragmented insulation to the wetwell. In the absence of more definitive information, the Regulatory Guide recommends that licensees assume a drywell debris transport fraction of 1.0. Accordingly, the Nuclear Regulatory Commission initiated research focused toward developing a technical basis to provide insights useful to regulatory oversight of licensee submittals associated with resolution of the postulated strainer blockage issue. Part of this program was directed towards experimental and analytical research leading to a more realistic specification of the debris transport through the drywell to the wetwell. To help focus this development into a cost effective effort, a panel, with broad based knowledge and experience, was formed to address the relative importance of the various phenomena that can be expected in plant response to postulated accidents that may produce strainer blockage. The resulting phenomena identification and ranking tables reported herein were used to help guide research. The phenomena occurring in boiling water reactors drywells was the specific focus of the panel, although supporting experimental data and calculations of debris transport fractions were considered.

  5. Upgrading instrumentation and control systems for plant safety and operation

    International Nuclear Information System (INIS)

    Martin, M.; Prehler, H.J.; Schramm, W.

    1997-01-01

    Upgrading the electrical systems and instrumentation and control systems has become increasingly more important in the past few years for nuclear power plants currently in operation. As the requirements to be met in terms of plant safety and availability have become more stringent in the past few years, Western plants built in the sixties and seventies have been the subject of manifold backfitting and upgrading measures in the past. In the meantime, however, various nuclear power plants are facing much more thorough upgrading phases because of the difficulties in obtaining spare parts for older equipment systems. As digital technology has become widespread in many areas because of its advantages, and as applications are continuously expanding, conventional equipment and systems are losing more and more ground as a consequence of decreasing demand. Merely because of the pronounced decline in demand for conventional electronic components it is possible for equipment manufacturers to guarantee spare parts deliveries for older systems only for specific future periods of time. In addition, one-off manufacture entails high costs in purchases of spare parts. As a consequence of current thinking more and more focusing on availability and economy, upgrading of electrical systems and instrumentation and control systems is becoming a more and more topical question, for older plants even to ensure completion of full service life. (orig.) [de

  6. Estimating collective dose in nuclear facilities, with emphasis on the design process

    International Nuclear Information System (INIS)

    Cohen, S.; Mann, B.

    1987-01-01

    The report presents a more accurate, systematic method than has been available previously for predicting worker doses which might be incurred during routine and non-routine work in radioactive areas. Besides assisting regulators with an analysis of the ''potential impact on radiological exposures of facility employees'' now required under the new backfit rule (10 CFR 50.109c), this predictive model will also help licensees conserve dollars as well as dose because it can be employed very early in the engineering design phase of a modification, when adjustments can still be made easily to change orders. Such early estimates make good business sense because they will facilitate planning, labor loading, costing, resource and equipment scheduling, and overall coordination of both single and repetitive projects. Also, with the support of corporate management, radiation protection coordinators can introduce the model into training programs to acquaint design engineers and others with dose calculation techniques. The importance assigned by nuclear industry senior management to the principle of ALARA and the reduction of collective worker dose is measured, in large part, by demonstrated efforts to integrate the control of radiation exposure fully into the overall planning function of nuclear facility management. That integration will be fostered through the use of this approach

  7. EPRI's zebra mussel monitoring and control guidelines

    International Nuclear Information System (INIS)

    Mussalli, Y.G.; Armor, A.; Edwards, R.; Mattice, J.; Miller, M.; Nott, B.; Tsou, J.L.

    1992-01-01

    The Electric Power Research Institute (EPRI) Zebra Mussel Monitoring and Control Guidelines is a comprehensive compilation of US and European practices. The zebra mussel has infested all the Great Lakes and is positioned to spread to the adjoining river basins. The impact of the zebra mussel on power plants is as a biofouler clogging water systems and heat exchangers. The EPRI guidelines discuss the distribution of the zebra mussel in the US, identification of the zebra mussel, potential threats to power plants, and methods to initiate the monitoring and control program. Both preventive and corrective measures are presented. Preventive measures include various monitoring methods to initiate control techniques. The control techniques include both chemical and nonchemical together with combining techniques. Corrective methods include operational considerations, chemical cleaning, and mechanical/physical cleaning. It may also be possible to incorporate design changes, such as open to closed-loop backfit, backflushing, or pretreatment for closed systems. Table 1 shows a matrix of the monitoring methods. Table 2 presents a control matrix related to nuclear, fossil, and hydro raw water systems. Table 3 is a summary of the applicability of treatments to the various raw water systems. Appendixes are included that contain specifications to aid utilities in implementing several of the control technologies

  8. Regulatory aspects of radiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Randall, P.N.

    1979-01-01

    One purpose of this conference, is to re-examine the conventional wisdom about neutron radiation embrittlement and the methods used to counteract embrittlement in reactor vessels. Perhaps, there have been sufficient advances in fracture mechanics, core physics, dosimetry, and physical metallurgy to permit a forward step in the quantitative treatment of the subject. Certainly this would be consistent with the position of the U.S. Nuclear Regulatory Commission (the NRC) in general. ''There has been a continued evolution toward increased specificity.'' This statement appeared in the response prepared by the staff to a request from the Commission to explain how the staff decides to apply a new requirement and to whom, i.e., to back-fit or forward-fit-only or whatever. Pressure for increased specificity, i.e., for fleshing out general design criteria, comes from ''technical surprises'' in the form of operating experiences or from research information, and from attempts to improve our confidence in the safety of plants, especially new plants. Our goal is to have anticipated and evaluated all possible modes of failure with sufficient quantitativeness that the probability of failure can be estimated with some accuracy. Failing this, regulators demand large margins of safety to cover our ignorance

  9. Handbook for quick cost estimates. A method for developing quick approximate estimates of costs for generic actions for nuclear power plants

    International Nuclear Information System (INIS)

    Ball, J.R.

    1986-04-01

    This document is a supplement to a ''Handbook for Cost Estimating'' (NUREG/CR-3971) and provides specific guidance for developing ''quick'' approximate estimates of the cost of implementing generic regulatory requirements for nuclear power plants. A method is presented for relating the known construction costs for new nuclear power plants (as contained in the Energy Economic Data Base) to the cost of performing similar work, on a back-fit basis, at existing plants. Cost factors are presented to account for variations in such important cost areas as construction labor productivity, engineering and quality assurance, replacement energy, reworking of existing features, and regional variations in the cost of materials and labor. Other cost categories addressed in this handbook include those for changes in plant operating personnel and plant documents, licensee costs, NRC costs, and costs for other government agencies. Data sheets, worksheets, and appropriate cost algorithms are included to guide the user through preparation of rough estimates. A sample estimate is prepared using the method and the estimating tools provided

  10. Regulatory analysis for resolution of Unresolved Safety Issue A-46, seismic qualification of equipment in operating plants

    International Nuclear Information System (INIS)

    Chang, T.Y.; Anderson, N.R.

    1987-02-01

    The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform required safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants must be reassessed to determine whether requalification is necessary. The objective of technical studies performed under Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring these plants to meet the criteria that are applied to new plants. This report presents the regulatory analysis for Unresolved Safety Issue (USI) A-46. It includes: Statement of the Problem; the Objective of USI A-46; a Summary of A-46 Tasks; a Proposed Implementation Procedure; a Value-Impact Analysis; Application of the Backfit Rule; 10 CFR 50.109; Implementation; and Operating Plants To Be Reviewed to USI A-46 Requirements

  11. Comparison of control systems applied to the handling of radioactive reactor components

    International Nuclear Information System (INIS)

    Robinson, C.; Harris, E.G.; Dyer, P.C.; Williams, J.G.B.

    1985-01-01

    The first generation of nuclear power stations have individual reactors each incorporating complete facilities for servicing components and refuelling. In the later designs, each power station has two reactors which are connected by a central block. This central block contains one set of facilities to service both reactors, but to improve the station capability, some of these are to be replicated. The central block incorporates a hoist well which was used during construction for the accessing of complete components. On completion of this work, the physical size of the hoist well is such as to permit the incorporation of additional facilities if these are shown to be operationally and economically desirable. Since a number of years of power operation has elapsed, the advantages of back-fitting to existing fuel-handling facilities has been illustrated. Since the mechanical arrangements and operating procedures are substantially similar for both the original and new handling facilities, the paper will illustrate the control systems provided for each. The configuration of the system is arranged to have two channels of control which complies with the current standard requirements in the United Kingdom. These requirements are more stringent than when the existing facility was designed and constructed, as described in the relevant sections of the paper. The new system has been designed and is being manufactured to comply with the Central Electricity Generating Board standard for nuclear fuel route interlock and control systems. (author)

  12. Organization and conduct of IAEA fire safety reviews at nuclear power plants

    International Nuclear Information System (INIS)

    1998-01-01

    The importance of fire safety in the safe and productive operation of nuclear power plants is recognized worldwide. Lessons learned from experience in nuclear power plants indicate that fire poses a real threat to nuclear safety and that its significance extends far beyond the scope of a conventional fire hazard. With a growing understanding of the close correlation between the fire hazard in nuclear power plants and nuclear safety, backfitting for fire safety has become necessary for a number of operating plants. However, it has been recognized that the expertise necessary for a systematic independent assessment of fire safety of a NPP may not always be available to a number of Member States. In order to assist in enhancing fire safety, the IAEA has already started to offer various services to Member States in the area of fire safety. At the request of a Member State, the IAEA may provide a team of experts to conduct fire safety reviews of varying scope to evaluate the adequacy of fire safety at a specific nuclear power plant during various phases such as construction, operation and decommissioning. The IAEA nuclear safety publications related to fire protection and fire safety form a common basis for these reviews. This report provides guidance for the experts involved in the organization and conduct of fire safety review services to ensure consistency and comprehensiveness of the reviews

  13. Basic national requirements for safe design, construction and operation

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1980-01-01

    Nuclear power plants have to be save. Vendors and utilities operating such plants, are convinced that their plants meet this requirement. Who, however, is establishing the safety requirements to be met by those manufacturing and operating nuclear power plants. What are the mechanisms to control whether the features provided assure the required safety level. Who controls whether the required and planned safety features are really provided. Who is eventually responsible for assuring safety after commissioning of a nuclear power plant. These fundamental questions being raised in many discussions on safety and environmental protection are dealt with in the following sections: (1) Fundamental safety requirements on nuclear power plants, in which such items as risk, legal bases and licensing procedure are discussed, (2) Surveillance during construction, in which safety analysis report, siting, safety evaluation, document examination, quality assurance, and commissioning testing are dealt with, (3) Operating tests and conditions in which recurrent inspections, environmental protection during operation, investigation of abnormal occurences and backfitting requirements as reviewed, and (4) Safety philosophy and safety policy to conclude this presentation. The German approach to nuclear safety serves as an example for an effective way of assuring safe nuclear power. (orig.)

  14. Analyzing the Impact of Residential Building Attributes, Demographic and Behavioral Factors on Natural Gas Usage

    Energy Technology Data Exchange (ETDEWEB)

    Livingston, Olga V.; Cort, Katherine A.

    2011-03-03

    This analysis examines the relationship between energy demand and residential building attributes, demographic characteristics, and behavioral variables using the U.S. Department of Energy’s Residential Energy Consumption Survey 2005 microdata. This study investigates the applicability of the smooth backfitting estimator to statistical analysis of residential energy consumption via nonparametric regression. The methodology utilized in the study extends nonparametric additive regression via local linear smooth backfitting to categorical variables. The conventional methods used for analyzing residential energy consumption are econometric modeling and engineering simulations. This study suggests an econometric approach that can be utilized in combination with simulation results. A common weakness of previously used econometric models is a very high likelihood that any suggested parametric relationships will be misspecified. Nonparametric modeling does not have this drawback. Its flexibility allows for uncovering more complex relationships between energy use and the explanatory variables than can possibly be achieved by parametric models. Traditionally, building simulation models overestimated the effects of energy efficiency measures when compared to actual "as-built" observed savings. While focusing on technical efficiency, they do not account for behavioral or market effects. The magnitude of behavioral or market effects may have a substantial influence on the final energy savings resulting from implementation of various energy conservation measures and programs. Moreover, variability in behavioral aspects and user characteristics appears to have a significant impact on total energy consumption. Inaccurate estimates of energy consumption and potential savings also impact investment decisions. The existing modeling literature, whether it relies on parametric specifications or engineering simulation, does not accommodate inclusion of a behavioral component. This

  15. Safeguarding the functioning of I and C systems

    International Nuclear Information System (INIS)

    Koehler, M.; Schoerner, O.

    1997-01-01

    On the basis of an analysis of instrumentation and control (I and C) systems in nuclear power plants the need is outlined to design digital instrumentation and control systems with forward looking technical features to serve both for plant operations management under normal conditions and for safety related problems in reactor and safety I and C. Siemens KWU not only took measures to safeguard the availability of existing permanently wired I and C systems, but also advanced the development, evaluation by experts, and commercialization of the Teleperm XS and Teleperm XP digital I and C systems. The working principle and the advantages of digital I and C systems are outlined briefly. A report is presented of the status of the licensing procedure and commercialization of the Teleperm XS safety I and C system. A number of examples are cited to explain the various possible uses of the systems discussed for new plants and for backfitting purposes, both in nuclear power plants by KWU and in facilities by other vendors. Siemens KWU remains a partner for in-plant and safety related I and C technology in nuclear power plants. This strategy is based on the principle of maintaining the availability of existing systems by adequate spare parts strategies as long as possible, and on the thorough innovation of I and C technology by the use of the forward looking Teleperm XS and Teleperm XP digital I and C systems. The Teleperm XS system is currently being introduced into the market and will be used on a broad basis in Germany and abroad in 1997/98. (orig.) [de

  16. Confirmation of soil radiation damping from test versus analysis

    International Nuclear Information System (INIS)

    Eidinger, J.M.; Mukhim, G.S.; Desmond, T.P.

    1987-01-01

    The work was performed to demonstrate that soil-structure interaction effects for nuclear plant structures can be accurately (and conservatively) predicted using the finite element or soil spring methods of soil-structure interaction analysis. Further, the work was done to investigate the relative importance of soil radiation versus soil material damping in the total soil damping analytical treatment. The analytical work was benchmarked with forced vibration tests of a concrete circular slab resting on the soil surface. The applied loading was in the form of a suddenly applied pulse load, or snapback. The measured responses of the slap represent the free vibration of the slab after the pulse load has been applied. This simplifies the interpretation of soil damping, by the use of the logarithmic decay formulation. To make comparisons with the test results, the damping data calculated from the analytical models is also based on the logarithmic decay formulation. An attempt is made to differentiate the observed damped behavior of the concrete slab as being caused by soil radiation versus soil material damping. It is concluded that both the traditional soil radiation and material damping analytical simplifications are validated by the observed responses. It is concluded that arbitrary 'conservative' assumptions traditionally made in nuclear plant soil-structure interaction analyses are indeed arbitrary, and not born out by physical evidence. The amount of conservatism introduced by limiting total soil damping to values like 5% to 10% can be large. For the test slab sizes investigated, total soil damping is about 25%. For full size nuclear plant foundations, total soil damping is commonly in the 35% to 70% range. The authors suggest that full soil damping values (the combined radiation and material damping) should be used in the design, backfit and margin assessment of nuclear plants. (orig./HP)

  17. Need for consent of a law extending the operating life of nuclear power plants

    International Nuclear Information System (INIS)

    Degenhart, Christoph

    2010-01-01

    The article deals with the question whether a law extending nuclear power plant life beyond the residual periods of time laid down in the law of April 22, 2002 requires consent of the Federal Council. The Atomic Energy Act needed the consent of the Federal Council pursuant to Article 87c, Basic Law, as its Section 24 determines that central functions of licensing and supervision be exercised by the federal states on behalf of the Federal Government. This has not changed after the current version of the norm. Increasing the residual quotas of electricity by amending Annex 3 of Sec.7, Para.1a, Atomic Energy Act, per se does not require consent. This is a substantive provision. Sec.24, Atomic Energy Act, does not need to be amended. The Federal Council, which consented to the original legislation, thus does not bear continued responsibility for the law. Every law must be treated as a separate entity in terms of legislative method. The Federal Council, with its first consent to the piece of legislation, ''approves'' this systemic shift. Renewed consent is required only in case of another systemic shift. This is the case when the provision about administrative responsibility takes on a very different meaning and impact no longer supported by the earlier consent. According to decisions by the Federal Constitutional Court, this expressly applies also to administration by commission. What is required is a comparison of administrative duties before and after entry into force of the amending law; mere quantitative shifts of administrative burdens do not cause a systemic shift. Whether the inclusion of backfitting obligations would be associated with regulations in administrative procedures has not been decided. In its ruling of May 4, 2010, the Federal Constitutional Court confirms that these do not require consent within the framework of Art.85 Para.1, Basic Law. (orig.)

  18. Advanced human-system interface design review guidelines

    International Nuclear Information System (INIS)

    O'Hara, J.M.

    1990-01-01

    Advanced, computer-based, human-system interface designs are emerging in nuclear power plant control rooms as a result of several factors. These include: (1) incorporation of new systems such as safety parameter display systems, (2) backfitting of current control rooms with new technologies when existing hardware is no longer supported by equipment vendors, and (3) development of advanced control room concepts. Control rooms of the future will be developed almost exclusively with advanced instrumentation and controls based upon digital technology. In addition, the control room operator will be interfacing with more intelligent systems which will be capable of providing information processing support to the operator. These developments may have significant implications for plant safety in that they will greatly affect the operator's role in the system as well as the ways in which he interacts with it. At present, however, the only guidance available to the Nuclear Regulatory Commission (NRC) for the review of control room-operator interfaces is NUREG-0700. It is a document which was written prior to these technological changes and is, therefore, tailored to the technologies used in traditional control rooms. Thus, the present guidance needs to be updated since it is inadequate to serve as the basis for NRC staff review of such advanced or hybrid control room designs. The objective of the project reported in this paper is to develop an Advanced Control Room Design Review Guideline suitable for use in performing human factors reviews of advanced operator interfaces. This guideline will take the form of a portable, interactive, computer-based document that may be conveniently used by an inspector in the field, as well as a text-based document

  19. MOX fuel design and development consideration

    International Nuclear Information System (INIS)

    Yamate, K.; Abeta, S.; Suzuki, K.; Doi, S.

    1997-01-01

    Pu thermal utilization in Japan will be realized in several plants in late 1990's, and will be expanded gradually. For this target, adequacy of methods for MOX fuel design, nuclear design, and safety analysis has been evaluated by the committee of competent authorities organized by government in advance of the licensing application. There is no big difference of physical properties and irradiation behaviors between MOX fuel and UO 2 fuel, because Pu content of MOX fuel for Pu thermal utilization is low. The fuel design code for UO 2 fuel will be applied with some modifications, taking into account of characteristic of MOX fuel. For nuclear design, new code system is to be applied to treat the heterogeneity in MOX fuel assembly and the neutron spectrum interaction with UO 2 fuel more accurately. For 1/3 MOX fueled core in three loop plant, it was confirmed that the fuel rod mechanical design could meet the design criteria, with slight reduction of initial back-fitting pressure, and with appropriate fuel loading patterns in the core to match power with UO 2 fuel. With the increase of MOX fuel fraction in the core, control rod worth and boron worth decrease. Compensating the decrease by adding control rod and utilizing enriched B-10 in safety injection system, 100% MOX fueled core could be possible. Up to 1/3 MOX fueled core in three loop plant, no such modifications of the plant is necessary. The fraction of MOX fuel in PWR is designed to less than 1/3 in the present program. In order to improve Pu thermal utilization in future, various R and D program on fuel design and nuclear design are being performed, such as the irradiation program of MOX fuel manufactured through new process to the extent of high burnup. (author). 8 refs, 9 figs, 2 tabs

  20. Initial Startup and Testing of the Fort St. Vrain HTGR - Lessons Learned which May Be Useful to the HTR-PM

    International Nuclear Information System (INIS)

    Brey, Larry H.

    2014-01-01

    Lessons Learned: Although the HTR-PM and FSV incorporate significant differences in their designs, there are lessons to be learned that are applicable to both plants. This is especially important for key systems that incorporate first-of-a-kind equipment. Basically, these lessons are just an application of common sense. • Complexity Breeds Unavailability. Incorporate system/components that are ruggedly simple in design with a history of reliable operation and minimal maintenance. • Assure Strong Expertise and Funding for this First HTR-PM. Quite likely, the successful startup and operation of this plant will require a level of support considerably greater than a typical nuclear plant. • Be Very Attentive to the Design Aspects of first-of-a-kind Components in the Class 1, Safety-Related Portions of the Plant. For example; a generic metallurgical failure could easily lead to a very long plant shutdown in order to redesign the failed component, re-license, manufacture, install and test prior to plant resuming plant operation. • Where Possible, Test all Key Systems/Components Prior to Installation using Actual Plant Configuration & Operating Characteristics This will help assure operational capability prior to application of nuclear heat. • Never Attempt to Start an Innovative Nuclear Power Plant Without First Having the Proper Regulatory Guides and Criteria in Place. FSV was licensed as a Research Facility. There was no Standard Review Plan or Regulatory Guides in place for the NRC (or PSC) to utilize in regulating this HTGR. • Do Not Be Reluctant to Incorporate a Generous Over-Build Capability into Systems/Components. It is significantly easier to design extra margin into the original compressors, pumps and motors than to be required to backfit into larger units after plant start-up. • Assure All Safety Documents Reflect the Actual Capability of the Plant to Respond to Accidents Described in the Safety Analysis. FSV was limited to 82% power during the

  1. An analysis of nuclear power plant operating costs

    International Nuclear Information System (INIS)

    1988-01-01

    This report presents the results of a statistical analysis of nonfuel operating costs for nuclear power plants. Most studies of the economic costs of nuclear power have focused on the rapid escalation in the cost of constructing a nuclear power plant. The present analysis found that there has also been substantial escalation in real (inflation-adjusted) nonfuel operating costs. It is important to determine the factors contributing to the escalation in operating costs, not only to understand what has occurred but also to gain insights about future trends in operating costs. There are two types of nonfuel operating costs. The first is routine operating and maintenance expenditures (O and M costs), and the second is large postoperational capital expenditures, or what is typically called ''capital additions.'' O and M costs consist mainly of expenditures on labor, and according to one recently completed study, the majoriy of employees at a nuclear power plant perform maintenance activities. It is generally thought that capital additions costs consist of large maintenance expenditures needed to keep the plants operational, and to make plant modifications (backfits) required by the Nuclear Regulatory Commission (NRC). Many discussions of nuclear power plant operating costs have not considered these capital additions costs, and a major finding of the present study is that these costs are substantial. The objective of this study was to determine why nonfuel operating costs have increased over the past decade. The statistical analysis examined a number of factors that have influenced the escalation in real nonfuel operating costs and these are discussed in this report. 4 figs, 19 tabs

  2. The Spanish nuclear adventure. From the past into the future

    International Nuclear Information System (INIS)

    Gonzalez de Ubieta, A.

    1986-10-01

    The paper describes the conditions under which nuclear power has been developed in Spain and the characteristics of such development. Electric utilities have been free to follow the laws of the market with the only restrictions imposed by safety regulations and national interest. The resulting programme is characterized by an early start, a stepwise introduction of plants, a diversity of types and suppliers, an important domestic contribution, an early consideration to fuel cycle and to administrative procedures. The paper analyzes the operating experience. For the three plants belonging to the first generation (1968-72) the load factor varies between 64% for the BWR and 74% for the GCR. Likewise the availability factor goes from 70% to 89% for the same units. In the LWR's important backfitting actions have been performed covering waste treatment, emergency core cooling and electric power supplies, among others. An important fraction of the recirculation system for the BWR has been replaced due to stress corrosion cracking. The operating experience for the second generation of nuclear power plants (1981-85) is also included, even though it is not considered representative. Load factors go from 49% for Almaraz one to 71% for Almaraz two. Availability factors range from 61% for Asco one to 80% for Almaraz two. The commissioning of these stations shows a rather large number of unscheduled shutdowns. The PWR's have experienced important modifications in their W D-3 steam generators. The 1983 National Energy Plan, reevaluated in 1985, has limited the installed nuclear power to 7,7 GWe by 1992 on the basis of an assumed excess capacity. This has forced the postponement of five units, four of them in an advanced stage of construction. Nevertheless nuclear power is still considered a solution for the intermediate and long terms. (author)

  3. The power of British Energy

    International Nuclear Information System (INIS)

    Hawley, R.

    1997-01-01

    When the power industry in Britain was privatized, British Energy plc (BE), whose head office is in Edingburgh, Scotland, was founded in July 1996. It is the only power utility in the world exclusively operating nuclear power stations. Operative business has remained the responsibility of the two regional supply companies, Nuclear Electric (NE) and Scottish Nuclear (SN) which, in addition to the modern PWR nuclear generating unit of Sizewell B, have included in the new holding company their advanced gas-cooled and gas-moderated reactor (AGR) units. The older gas-graphite reactor (GGR) plants were combined in the new Magnox Electric plc, Berkeley; at some later date, this company is to be merged with another nuclear power plant operator, British Nuclear Fuels plc (BNFL). Sizewell B, which was commissioned in 1995, is the last nuclear generating unit to be started up in the United Kingdom, for the time being. In times of low raw material prices and the need for a quick return on invested capital, BE is reluctant to run the risk associated with tying up capital for a long time. Instead, the company has backfitted its plants so that the production of electricity from nuclear power in Britain in 1996 of 92,476 GWh was increased by almost 10% over the 1995 level of 84,174 GWh. In addition to modernization and rationalization at home, BE together with Sizewell B vendor Westinghouse is engaged worldwide in the development and commercialization of future advanced reactors. This ensures that the know-how accumulated will be preserved and will be available for new nuclear power plants to be built in Britain in the next century. (orig.)

  4. Lessons learned from full-scale vibration tests on nuclear power plant auxiliary structure in Switzerland

    International Nuclear Information System (INIS)

    Berger, E.; Tinic, S.

    1988-01-01

    The Beznau Nuclear Power Plant is located in northern Switzerland. The plant is owned and operated by the Nordostschweizerische Kraftwerke AG (NOK) in Baden, Switzerland. It is a twin unit plant (2 x 350 MWe) which was designed in the early 1960's and placed into commercial operation between 1969 and 1971. In connection with a major backfit project, which will improve the safety of the plant against external events, the free-standing boric water tanks had to be relocated and were replaced by two boric water tanks in a new building (the so called BOTA-building). It enabled to plan and perform full scale vibration tests.The scope of experimental investigation was to determine the eigenfrequencies and damping values for fundamental soil-structure interaction. The vibration tests allowed identification of the important modes of the soil-structure system in the range 3 to 15 Hz. The excitation was strung enough to generate accelerations in the structure comparable to those of a small earthquake. From the comparisons of computed and measured results it is concluded that the rocking frequency can be reasonably well predicted by either Finite Element or Lumped Parameter models with springs simulating the soil-foundation stiffness, provided in the case of the latter the embedment is taken into account. The prediction of the amplitude of structural response appears to be more difficult, as shown by the differences in the mode shapes. In the frequency range 8 to 10 Hz the agreement between computed and test results was less satisfactory. The actual structural behaviour turned out to be more complex than expected and needs further investigation with the aid of more refined models for the soil-structure system

  5. Use of remote visual in-service inspection on nuclear power plants of the CEGB

    International Nuclear Information System (INIS)

    James, D.W.

    1985-01-01

    The main responsibility of the Remote Inspection Group is the design, development and procurement of the remote visual inspection equipment provided by the Generation Development and Construction Division as part of the extent of the supply for all the Central Electricity Generating Board's (CEGB) advanced gas-cooled reactors (AGR). The paper describes the operation of this equipment, together with the low light-level TV cameras that have been developed for carrying out routine remote visual inspections. The camera, known as the television remote inspection unit multi-purpose head (TRIUMPH), has been designed as a series of modules. With this system it is possible to take advantage of improvements in a particular part of the camera system and to arrange to backfit an improved module to existing TRIUMPHs. To minimize the time for carrying out routine inspections during shutdown, the AGRs have been provided with storage training and test facilities. These facilities are provided with full size mock-ups of the reactor internals so that the inspection equipment can be tested and the operating staff trained before the equipment is used on the reactor. One of the other responsibilities of the Remote Inspection Group is to carry out specific power plant remote visual inspections which are required to minimize costly plant shutdowns and construction delays. Examples are given of successful inspections that have been carried out. Over 12 years' experience has now been obtained in carrying out, at short notice, difficult inspections which involve tortuous access routes. The CEGB now holds a wide range of fibrescope and small TV cameras, together with the equipment for placing the viewing device in the correct location. A number of special fibrescopes have been developed for specific inspection needs and details of these, together with other fibrescopes owned by the CEGB, are provided. (author)

  6. Development of default uncertainties for the value/benefit attributes in the regulatory analysis technical evaluation handbook

    International Nuclear Information System (INIS)

    Gallucci, Raymond H.V.

    2016-01-01

    Highlights: • Uncertainties for values/benefits. • Upper bound four times higher than mean. • Distributional histograms. - Abstract: NUREG/BR-0184, Regulatory Analysis Technical Evaluation (RATE) Handbook, was produced in 1997 as an update to the original NUREG/CR-3568, A Handbook for Value-Impact Assessment (1983). Both documents, especially the later RATE Handbook, have been used extensively by the USNRC and its contractors not only for regulatory analyses to support backfit considerations but also for similar applications, such as Severe Accident Management Alternative (SAMA) analyses as part of license renewals. While both provided high-level guidance on the performance of uncertainty analyses for the various value/benefit attributes, detailed quantification was not of prime interest at the times of the Handbooks’ development, defaulting only to best estimates with low and high bounds on these attributes. As the USNRC examines the possibility of updating the RATE Handbook, renewed interest in a more quantitative approach to uncertainty analyses for the attributes has surfaced. As the result of an effort to enhance the RATE Handbook to permit at least default uncertainty analyses for the value/benefit attributes, it has proven feasible to assign default uncertainties in terms of 95th %ile upper bounds (and absolute lower bounds) on the five dominant value/benefit attributes, and their sum, when performing a regulatory analysis via the RATE Handbook. Appropriate default lower bounds of zero (no value/benefit) and an upper bound (95th %ile) that is four times higher than the mean (for individual value/benefit attributes) or three times higher (for their summation) can be recommended. Distributions in the form of histograms on the summed value/benefit attributes are also provided which could be combined, after appropriate scaling and most likely via simulation, with their counterpart(s) from the impact/cost analysis to yield a final distribution on the net

  7. Designing for nuclear power plant maintainability and operability

    International Nuclear Information System (INIS)

    Pedersen, T.J.

    1998-01-01

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

  8. US nuclear safety review and experience

    International Nuclear Information System (INIS)

    Gilinsky, V.

    1977-01-01

    The nuclear safety review of commercial nuclear power reactors has changed over the years from the relatively simple review of Dresden 1 in 1955 to the highly complex and sophisticated regulatory process which characterizes today's reviews. Four factors have influenced this evolution: (1) maturing of the technology and industry; (2) development of the regulatory process and associated staff; (3) feedback of operating experience; and (4) public awareness and participation. The NRC's safety review responsibilities start before an application is tendered and end when the plant is decommissioned. The safety review for reactor licensing is a comprehensive, two-phase process designed to assure that all the established conservative acceptance criteria are satisfied. Operational safety is assured through a strong inspection and enforcement program which includes shutting down operating facilities when necessary to protect the health and safety of the public. The safety of operating reactors is further insured through close regulation of license changes and selective backfitting of new regulatory requirements. An effective NRC standards development program has been implemented and coordinates closely with the national standards program. A confirmatory safety research program has been developed. Both of these efforts are invaluable to the nuclear safety review because they provide the staff with key tools needed to carry out its regulatory responsibilities. Both have been given increased emphasis since the formation of the NRC in 1975. The safety review process will continue to evolve, but changes will be slower and more deliberate. It will be influenced by standardization, early site reviews and development of advanced reactor concepts. New legislation may make possible changes which will simplify and shorten the regulatory process. Certainly the experience provided by the increasing number and types of operating plants will have a very strong impact on future trends in the

  9. Paul Scherrer Institut Scientific Report 2001. Volume IV: Nuclear Energy and Safety

    International Nuclear Information System (INIS)

    Smith, B. L.

    2002-03-01

    The year 2001 was marked by the extensive audit of energy research at PSI. The audit took place on 21/22 June, and nuclear energy research was evaluated by five international experts. It was possible to present the quality and relevance of our research in a convincing way. This, together with optimal use of the available resources in our department Nuclear Energy and Safety (NES), prompted the highly welcome result that the auditors attested a high degree of scientific quality to the NES projects in general, with some of them judged to be of world-class standard. They recommended all audited projects be maintained, adequate resources to be allocated accordingly and, if necessary, an increase of public funding in nuclear energy research to be granted to achieve this. Finally, it was recommended that energy research should be explicitly mentioned in the statement of the ultimate mission of PSI. At the level of 'daily work', efforts in several directions related to the future of the Hotlab were one of the main focuses during the past year. On the one hand, the safety-related backfitting of the facility (particularly in regard to fire protection) has been continued, and was coupled with intensive planning studies, and with a thorough radiological cleaning of the labs; this invariably led to some impediment to the current research activities. Despite external burdens, which have led to some delays and additional costs, three of the four refurbishment phases have been completed, and the corresponding laboratories were brought back into operation after inspection and approval by the regulatory authority HSK. Given the size and complexity of the project, progress must be considered very satisfactory. A list of scientific publications in 2000 is also provided

  10. Paul Scherrer Institut Scientific Report 2001. Volume IV: Nuclear Energy and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Smith, B.L. (ed.)

    2002-03-01

    The year 2001 was marked by the extensive audit of energy research at PSI. The audit took place on 21/22 June, and nuclear energy research was evaluated by five international experts. It was possible to present the quality and relevance of our research in a convincing way. This, together with optimal use of the available resources in our department Nuclear Energy and Safety (NES), prompted the highly welcome result that the auditors attested a high degree of scientific quality to the NES projects in general, with some of them judged to be of world-class standard. They recommended all audited projects be maintained, adequate resources to be allocated accordingly and, if necessary, an increase of public funding in nuclear energy research to be granted to achieve this. Finally, it was recommended that energy research should be explicitly mentioned in the statement of the ultimate mission of PSI. At the level of 'daily work', efforts in several directions related to the future of the Hotlab were one of the main focuses during the past year. On the one hand, the safety-related backfitting of the facility (particularly in regard to fire protection) has been continued, and was coupled with intensive planning studies, and with a thorough radiological cleaning of the labs; this invariably led to some impediment to the current research activities. Despite external burdens, which have led to some delays and additional costs, three of the four refurbishment phases have been completed, and the corresponding laboratories were brought back into operation after inspection and approval by the regulatory authority HSK. Given the size and complexity of the project, progress must be considered very satisfactory. A list of scientific publications in 2000 is also provided.

  11. Development of a technical process concerning the immobilisation of nuclear waste by embedding into ceramic matrix

    International Nuclear Information System (INIS)

    Schubert, G.; Krause, H.

    1993-12-01

    Ceramic is considered a highly qualified matrix for the embedding of all radioactive waste concentrates arising from reprocessing and fabricating UO 2 /PuO 2 -mixed oxide fuel elements and it may take up all long-lived or highly active radionuclides. Parallel to product development a technically feasible process has been started. The wastes are mixed with the ceramics-forming agents in a wet medium. A double-shaft extruder may be used. Backfitting of the extruder for use in a hot cell may be carried out easily. Experiments are presented and conceptions developed as to how the facility may be designed under aggravated boundary conditions of irradiation and remote handling. The process consists of the following stages: Preliminary treatment of the four waste suspensions, without dehydration; continuous dosage into a double-shaft extruder, where preliminary drying and then addition of the fifth waste type (dry ash) as well as of the mixture of ceramics-forming agents takes place; mixing and preferably extrusion. Heat treatment from the drying and calcination temperatures up to the sintering temperature of 1250-1300 C in a stationary heated electric furnace, filling of the hot material into canisters, filling of the cavities with liquid glas, and sealing of the cansiters. Except for an experiment with dissolver residues, all experiments were inactive. Conventional devices were applied with the aim of investigated their suitability for the process as well as for the conditions of remote handling and inrradiation. A facility, which was to be located downstream of a 350 t/a reprocessing plant, would have to have a throughput of about 40 kg/h ceramic product or 6 canisters per day. (orig./HP) [de

  12. If there had been new regulatory standards, could the Fukushima accident have been avoided?

    International Nuclear Information System (INIS)

    Hashizume, Hidetoshi; Aoki, Takayuki

    2015-01-01

    The Japan Society of Maintenology, at its Tohoku-Hokkaido Branch, established 'Study group for virtual back-fit simulation,' and conducted a simulation to clarify the effects of tsunami countermeasures after the Fukushima Nuclear Accident. These tests were carried out at Onagawa Unit 2 plant under the presence or absence of safety measures, and tsunami height of 23.8 m (3.11 Tsunami height 13.8 m + 10 m), 29 m, and 34 m. As a result of the study, when tsunami height was 23.8 m, under the condition without earthquake measures as before the 3.11 Earthquake, safety-related equipment was submerged, leading to core damage. Under the condition without a seawall but with the safety measures after the 3.11 Earthquake, although safety equipment was submerged, safety measures could effectively work leading to a safe cold shutdown. Under the conditions of seawalls + safety measures, flooding to the building block did not occur, and there was no effect on safety equipment. Under the conditions of seawalls + safety measures + tsunami height 29 m, there was not the effect as seen above. Under conditions of seawalls + safety measures + tsunami height 34 m, although outdoor safety system equipment was submerged, cold shutdown was achieved through the safety measures. In this way, safety measures after the 3.11 Earthquake can significantly improve safety. It is recommended for electric power enterprises to carry out such investigation/evaluation, including for incidents other than tsunami. (A.O.)

  13. PSA in America

    International Nuclear Information System (INIS)

    Linn, M.A.; Cunningham, M.A.; Johnson, D.H.

    1996-01-01

    Although the concept of acceptable risk has always been the foundation of the nuclear industry design, the use of formal PSA (or PRA-probabilistic risk assessment) in the U.S. nuclear power industry has followed an unusual path in arriving at its current level of notability. Prior to 1975, probabilistic evaluations were limited to a few specific applications such as the evaluation of man-made (i.e., airplane crashes) and natural (i.e., earthquakes) hazards. In 1975, the industry was introduced to comprehensive PSA by the Reactor Safety Study (WASH-1400). However, the study languished in relative obscurity until the accident at Three Mile Island 2 (TMI-2) in 1979. This event significantly altered the industry's view of severe accidents in the U.S. and worldwide. Investigative committees of TMI-2 recommended that PSA techniques be more widely used to augment the traditional deterministic methods of determining nuclear plant safety. This initiated an unprecedented effort by nuclear regulators and licensees worldwide to significantly improve the state of knowledge of severe accidents at nuclear power plants. In the U.S., use of PSA began to increase as evidenced by its application in the anticipated transient without scram and station blackout rulemakings, generic issue prioritization and resolution, risk-based inspection guidelines, backfit policy, and technical specification improvements. However, broad application of probabilistic techniques to the industry as a whole was initiated in 1986 with the publication of Safety Goals for the Operation of Nuclear Power Plant; Policy Statement. This put PSA front and center in the U.S. regulatory arena by open-quotes establish[ing] goals that broadly define an acceptable level of radiological risk that might be imposed on the public as a result of nuclear power plant operation.close quotes Both qualitative safety goals and quantitative objectives were articulated in this policy statement

  14. Safety assessment and regulatory strategy for NPP I and C modernization projects

    International Nuclear Information System (INIS)

    Manners, S.; Blocquel, Ch.

    1999-10-01

    IPSN is the technical support for the French nuclear safety authority (DSIN), but also acts independently. Through our participation at this IAEA meeting we wish to further our appreciation of the industry position for I and C modernization projects. We will present some of the concerns of the safety assessor and safety authority for such projects. We hope to share our experiences and views concerning current strategies for I and C modernization and licensing from. In our experience with NPP I and C programmes, the need for modification is most often not directly linked to safety. For our safety assessment we have to identify clearly and, as far as possible, categorize the safety relevance of the specified modifications and all safety impact in its implementation. Modernization can be simply for reasons of replacement of obsolete existing equipment or it can be linked to functional evolutions; safety functions may be directly or indirectly affected. The state of the art I and C solutions proposed by today's modernization programs have many benefits, but also pose a certain number of difficulties for the safety demonstration. On the implementation side, the safety assessment for a modernization project has to take into consideration specific issues compared with that for new plant. These include interface and compatibility with the existing installation, issues relating to 'on line' installation and commissioning, as well as operational issues concerning the changeover and trail periods. A further subject for discussion concerns how our regulatory requirements apply to modernization. We must as a minima comply with the requirements of the period. To what measure must we apply current or future (under development or for future reactor designs) standards? How can we tie in with requirements and legislation for new projects? Do we make a special case for back-fits? (authors)

  15. Accident localization system with jet condensers for VVER 440-V 230 NPP at Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Murani, J.

    1995-01-01

    The operational safety of the V1 nuclear power plant (NPP) is unsatisfactory and does not correspond to present requirements as to nuclear safety. Further NPP operation after 1995 is conditional on nuclear safety enhancement to a level comparable with that in West European countries. This aim should be achieved by a principal reconstruction involving in addition to others also backfitting the V1 NPP with technical facilities aimed at coping with a design basis accident (DBA).To cope with such an accident the Power Equipment Research Institute (VUEZ) designed an accident localization system with jet condensers. This system consists of (a) an air trap (one for each unit, mutually interconnected) with an expansion bell enclosed within, placed on a plate with 200 pipes of jet condensers passing through, and (b) a connecting duct between the hermetic zone and the air trap. The vertical jet condenser is an essential element of the system designed for steam condensation. Apart from condensation it serves as a water seal separating units 1 and 2.Demonstration tests of the jet condenser (model 1:1) condensing function were carried out at the testing unit of the All-Union Research Institute for NPP Operation (VNIIAES), Moscow in Kashir, 11-22 September 1992. These experiments proved the jet condenser ability to ensure complete condensation of the steam produced. Experimental verification of the sealing function (model 1:1) was carried out at the testing unit of the VUEZ Tlmace. These experiments concerning the dynamics and overpressure in the free space above the pool were close to the conditions in the air trap during DBA. The jet condenser height was proved to be sufficient to ensure the sealing function. Design and experimental work has been implemented in close cooperation with Russian experts Mr. V.N. Bulynin from the VNIIAES, Moscow, and Mr. M.V. Kuznecov from the Scientific and Engineering Center for Nuclear and Radiological Safety, Moscow. (orig.)

  16. Absorber materials in CANDU PHWR's

    International Nuclear Information System (INIS)

    Price, E.G.; Boss, C.R.; Novak, W.Z.; Fong, R.W.L.

    1995-03-01

    In a CANDU reactor the fuel channels are arranged on a square lattice in a calandria filled with heavy water moderator. This arrangement allows five types of tubular neutron absorber devices to be located in a relatively benign environment of low pressure, low temperature heavy water between neighbouring rows of columns of fuel channels. This paper will describe the roles of the devices and outline the design requirements of the absorber component from a reactor physics viewpoint. Nuclear heating and activation problems associated with the different absorbers will be briefly discussed. The design and manufacture of the devices will be also discussed. The control rod absorbers and shut off materials are cadmium and stainless steel. In the tubular arrangement, the cadmium is sandwiched between stainless steel tubes. This type of device has functioned well, but there is now concern over the availability and expense of cadmium which is used in two types of CANDU control devices. There are also concerns about the toxicity of cadmium during the fabrication of the absorbers. These concerns are prompting AECL to study alternatives. To minimize design changes, pure boron-10 alloyed in stainless steel is a favoured option. Work is underway to confirm the suitability of the boron-loaded steel and identify other encapsulated absorber materials for practical application. Because the reactivity devices or their guide tubes span the calandria vessel, the long slender components must be sufficiently rigid to resist operational vibration and also be seismically stable. Some of these components are made of Zircaloy to minimize neutron absorption. Slow irradiation growth and creep can reduce the spring tension, and periodic adjustments to the springs are required. Experience with the control absorber devices has generally been good. In one instance liquid zone controllers had a problem of vibration induced fretting but a designed back-fit resolved the problem. (author). 3 refs., 1

  17. LBB considerations for a new plant design

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Mandava, P.R.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1997-04-01

    The leak-before-break (LBB) methodology is accepted as a technically justifiable approach for eliminating postulation of Double-Ended Guillotine Breaks (DEGB) in high energy piping systems. This is the result of extensive research, development, and rigorous evaluations by the NRC and the commercial nuclear power industry since the early 1970s. The DEGB postulation is responsible for the many hundreds of pipe whip restraints and jet shields found in commercial nuclear plants. These restraints and jet shields not only cost many millions of dollars, but also cause plant congestion leading to reduced reliability in inservice inspection and increased man-rem exposure. While use of leak-before-break technology saved hundreds of millions of dollars in backfit costs to many operating Westinghouse plants, value-impacts resulting from the application of this technology for future plants are greater on a per plant basis. These benefits will be highlighted in this paper. The LBB technology has been applied extensively to high energy piping systems in operating plants. However, there are differences between the application of LBB technology to an operating plant and to a new plant design. In this paper an approach is proposed which is suitable for application of LBB to a new plant design such as the Westinghouse AP600. The approach is based on generating Bounding Analyses Curves (BAC) for the candidate piping systems. The general methodology and criteria used for developing the BACs are based on modified GDC-4 and Standard Review Plan (SRP) 3.6.3. The BAC allows advance evaluation of the piping system from the LBB standpoint thereby assuring LBB conformance for the piping system. The piping designer can use the results of the BACs to determine acceptability of design loads and make modifications (in terms of piping layout and support configurations) as necessary at the design stage to assure LBB for the, piping systems under consideration.

  18. International status of application of probabilistic risk analysis

    International Nuclear Information System (INIS)

    Cullingford, M.C.

    1984-01-01

    Probabilistic Risk Assessment (PRA) having been practised for about ten years and with more than twenty studies completed has reached a level of maturity such that the insights and other products derived from specific studies may be assessed. The first full-scale PRA studies were designed to develop the methodology and assess the overall risk from nuclear power. At present PRA is performed mostly for individual plants to identify core damage accident sequences and significant contributors to such sequences. More than 25 countries are utilizing insights from PRA, some from full-scale PRA studies and other countries by performing reliability analyses on safety systems identified as important contributors to one or more core melt sequences. Many Member States of the IAEA fall into one of three groups: those having (a) a large, (b) a medium number of reactor-years of operating experience and (c) those countries in the planning or feasibility study stages of a nuclear power programme. Of the many potential uses of PRA the decision areas of safety improvement by backfitting, development of operating procedures and as the basis of standards are felt to be important by countries of all three groups. The use of PRA in showing compliance with safety goals and for plant availability studies is held to be important only by those countries which have operating experience. The evolution of the PRA methodology has led to increased attention to quantification of uncertainties both in the probabilities and consequences. Although many products from performing a PRA do not rely upon overall risk numbers, increasing emphasis is being placed on the interpretation of uncertainties in risk numbers for use in decisions. International co-operation through exchange of information regarding experience with PRA methodology and its application to nuclear safety decisions will greatly enhance the widespread use of PRA. (author)

  19. Improving nuclear regulation. NEA regulatory guidance booklets volumes 1-14

    International Nuclear Information System (INIS)

    2011-01-01

    A common theme throughout the series of NEA regulatory guidance reports, or 'green booklets', is the premise that the fundamental objective of all nuclear safety regulatory bodies is to ensure that nuclear facilities are continuously maintained and operated in an acceptably safe manner. In meeting this objective the regulator must bear in mind that it is the operator that has responsibility for safely operating the nuclear facility; the role of the regulator is to assess and to provide assurance regarding the operator's activities in terms of assuming that responsibility. The full series of these reports was brought together in one edition for the first time in 2009 and was widely found to be a useful resource. This second edition comprises 14 volumes, including the latest on The Nuclear Regulator's Role in Assessing Licensee Oversight of Vendor and Other Contracted Services. The reports address various challenges that could apply throughout the lifetime of a nuclear facility, including design, siting, manufacturing, construction, commissioning, operation, maintenance and decommissioning. The compilation is intended to serve as a knowledge management tool both for current regulators and the new nuclear professionals and organisations entering the regulatory field. Contents: Executive Summary; Regulatory Challenges: 1. The Role of the Nuclear Regulator in Promoting and Evaluating Safety Culture; 2. Regulatory Response Strategies for Safety Culture Problems; 3. Nuclear Regulatory Challenges Related to Human Performance; 4. Regulatory Challenges in Using Nuclear Operating Experience; 5. Nuclear Regulatory Review of Licensee Self-assessment (LSA); 6. Nuclear Regulatory Challenges Arising from Competition in Electricity Markets; 7. The Nuclear Regulatory Challenge of Judging Safety Back-fits; 8. The Regulatory Challenges of Decommissioning Nuclear Reactors; 9. The Nuclear Regulator's Role in Assessing Licensee Oversight of Vendor and Other Contracted Services

  20. The safety of Ontario's nuclear power reactors. A scientific and technical review. Ontario Hydro Submission to the Ontario Nuclear Safety Review

    International Nuclear Information System (INIS)

    1987-01-01

    Ontario Hydro is responsible for the safety of its nuclear stations: safety analysis, design and construction, training of operators, operating practices, and maintenance procedures. The utility must demonstrate to the regulatory body and the public that it is capable of operating nuclear stations safely. the dedicated attention of management and workers alike has been given to the achievement of an excellent safety record. Safety begins with well understood corporate goals, objectives and policies, and the clear assignment of responsibilities to well-trained, competent people who have the relevant experience and the right information and equipment. A prime cause of both the Chernobyl and the Three Mile Island accidents was a breakdown in operational procedures and human factors. On the contrary, the pressure tube failure at Pickering unit 2 in 1983 was understood almost immediately by the operators, who took the correct steps to shut down the reactor. This success is related to well-designed control room information systems and good understanding of fundamentals by the operators. Increasingly, in the design of nuclear plant control and instrumentation systems and in training in Ontario Hydro, the well-being, capabilities and limitations of humans are being taken into account. This report describes the series of barriers between the radioactive material in the fuel and the series of barriers between the radioactive material in the fuel and the environment, and the stringent quality control and technical measures taken to make the likelihood of malfunctions very small. Defence in depth protection for the public is a feature of all Ontario Hydro nuclear stations. As safety-related systems are updated in new stations, improvements are in some cases being backfitted to older stations

  1. Leibstadt nuclear power station (KKL): The Future after twenty years of operation

    International Nuclear Information System (INIS)

    Schoenenberger, M.

    2005-01-01

    Switzerland's largest power plant, KKL (1 165 MW BWR), is situated on the Swiss side of the Rhine River, not far from the entry of the Aare River. In 2003, the plant generated some 17% of the electricity consumed in the country. In line with the importance of the plant, it shareholders are all major Swiss power utilities. KKL was connected to the power grid in December 1984. Its construction cost amounted to approx. euro 3 200 million. After some backfitting measures at an expense of approx. euro 200 million, the plant is now in excellent technical shape. Generating costs, which were very high in the beginning, have been greatly reduced in the meantime. This was helped by a decrease of borrowed capital, the favorable development of interest rates and, above all, the rise in annual production. This, in turn, was achieved in various programs increasing plant power, and by shortening the annual revision outages. From the coming year onward, costs could be below Eurocent 3.3/kWh. Also the organization and the staff of te plant are prepared for the future. They have demonstrated their fitness in various national and international reviews. Also the political environment is favorable, by and large. In 2003, the Swiss voting population so clearly rejected the two opt-out initiatives that there has been a lasting positive change in the policy of continuing the operation of existing plants. Also for KKL, the waste management problem is still unsolved. This is due primarily to political reasons. The envisaged repository for low-level waste was rejected in a referendum in 2003. Technically and in its organization, the Leibstadt Nuclear Power Station is ready for the future. The electricity generated at Leibstadt is desired and accepted politically. (orig.)

  2. Finnish experiences on licensing and using of programmable digital systems in nuclear power plants

    International Nuclear Information System (INIS)

    Haapanen, P.; Maskuniitty, M.; Heimburger, H.; Hall, L.E.; Manninen, T.

    1993-01-01

    Finnish utility companies, Imatran Voima Oy (IVO) and Teollisuuden Voima (TVO), and the licensing authority, the Finnish Centre for Radiation and Nuclear Safety (STUK), are preparing for a new nuclear power plant in Finland. Plant vendors are proposing programmable digital automation systems for both the safety-related and the operational I and C (instrumentation and control) systems in this new unit. Also in existing plant units the replacement of certain old analog systems with state-of-the-art digital ones will become necessary in the years to come. Licensing of programmable systems for safety critical applications requires a new approach due to the special properties and failure modes of these systems. The major difficulties seem to be in the assessment and quantification of software reliability. The Technical Research Centre of Finland has in co-operation with the authority and the utilities conducted a project (AJA) to develop domestically applicable licensing requirements, guidelines and practices. International standards, guidelines and licensing practices have been analyzed in order to specify national licensing requirements. The paper describes and discusses the findings and experiences of the AJA project so far. The experience in introducing advanced programmable digital control and computer systems in the operating nuclear power plants will be covered briefly. Although these systems are not safety-related but systems of more general interest regarding nuclear safety, some routines regarding the licensing of safety- related systems have been followed. In these backfitting and replacement projects some experience have been gained in how to license safety-related programmable systems. (Author) 31 refs., 2 figs

  3. 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)

  4. Experience in modernization of safety I and C in VVER 440 nuclear power plants Bohunice V1 and Paks

    International Nuclear Information System (INIS)

    Martin, M.

    2000-01-01

    For nuclear power plants which have been in operation for more than 15 years, backfitting or even complete replacement of the instrumentation and control (I and C) equipment becomes an increasingly attractive option, motivated not only by the objective to reduce the cost of I and C system maintenance and repair but also to prolong or at least to safeguard the plant life-time: optimized spare-part management through use of standard equipment; reduction of number and variety of different items of equipment by implementing functions stepwise in application software; adding new functionality in the application software which was not possible in the old technology due to lack of space; safeguarding of long-term After-Sales-Service. Some years ago Bohunice V1 NPP, Slovak Republic and Paks NPP, Hungary intended to replace parts of their Safety I and C, mainly the Reactor Trip System, the Reactor Limitation System and the Neutron Flux Excore Instrumentation and Monitoring Systems. After a Basic Engineering Phase in Bohunice V1 and a Feasibility Study in Paks both plants decided to use the Siemens' Digital Safety I and C System TELEPERM XS to modernize their plants. Both Bohunice, Unit 2 and Paks, Unit 1 have been back on line for over six months with the new Digital Safety I and C. At the present time Bohunice, Unit 1 and within the next few months Paks, Unit 2 will be replaced. Trouble-free start-ups and no major problems under operation in the first two plants were based on: thorough understanding of the VVER 440 technology; comprehensive planning together with the plant operators and authorities; the possibility to adapt TELEPERM XS to every plant type; the execution of extensive pre-operational tests. Regarding these modernization measures Siemens, as well as the above Operators, have gained considerable experience in the field of I and C Modernization in VVER 440 NPPs. Important aspects of this experience are: how to transfer the VVER technology to TELEPERM XS; how to

  5. Studies on the effectiveness of measures to maintain the integrity of pressurized components in German nuclear power plants. Final report

    International Nuclear Information System (INIS)

    Elmas, M.; Jendrich, U.; Michel, F.; Reck, H.; Schimpfke, T.; Walter, M.; Wenke, R.

    2013-03-01

    The overall objective of the project was to investigate the effectiveness of measures to maintain the as-built quality of the pressure-retaining components in German nuclear power plants. In particular, investigations were performed on the application of the break preclusion concept, existing monitoring systems and the significance of the pressure test as part of the inspection concept. Moreover, the KompInt knowledge base has been updated. Break preclusion for pipes was applied in all German plants already during planning or after commissioning to a varying extent. The basic features of the required assessments were considered in the German nuclear regulations for the first time by inclusion in the safety requirements for nuclear power plants of 2012. The requirements for assessments, differing in their degree of detail, in the interpretations of these safety requirements and in the safety standard KTA 3206 are still in the draft stage. For the first time, the vessels as well as housings of valves and pumps are also included in the concept. Through the use of advanced monitoring systems it was possible in German plants at an early stage to establish modes of operation that minimise the load on components, to carry out appropriate technical backfitting measures, and to identify damages. In plant areas where local water chemistry parameters may result that deviate from the specification, the effectiveness of water chemistry monitoring is limited. In this case, other operational measures must be taken. The results of the simulations performed with the help of the GRS-developed PROST computer code to determine the significance of pressure tests lead - in accordance with the results of operating experience evaluation - to the conclusion that pressure tests carried out within the pressure-retaining boundary contribute to safeguarding the integrity. The user-friendliness of the KompInt knowledge base has been increased by changing over to a new hardware, a software

  6. Thermomechanical behaviour of two heterogeneous tungsten materials via 2D and 3D image-based FEM

    International Nuclear Information System (INIS)

    Zivelonghi, Alessandro

    2011-01-01

    An advanced numerical procedure based on imaging of the material microstructure (Image- Based Finite Element Method or Image-Based FEM) was extended and applied to model the thermomechanical behaviour of novel materials for fusion applications. Two tungsten based heterogeneous materials with different random morphologies have been chosen as challenging case studies: (1) a two-phase mixed ductile-brittle W/CuCr1Zr composite and (2) vacuum plasma-sprayed tungsten (VPS-W 75 vol.%), a porous coating system with complex dual-scale microstructure. Both materials are designed for the future fusion reactor DEMO: W/CuCr1Zr as main constituent of a layered functionally graded joint between plasma-facing armor and heat sink whereas VPS-W for covering the first wall of the reactor vessel in direct contact with the plasma. The primary focus of this work was to investigate the mesoscopic material behaviour and the linkage to the macroscopic response in modeling failure and heat-transfer. Particular care was taken in validating and integrating simulation findings with experimental inputs. The solution of the local thermomechanical behaviour directly on the real material microstructure enabled meaningful insights into the complex failure mechanism of both materials. For W/CuCr1Zr full macroscopic stress-strain curves including the softening and failure part could be simulated and compared with experimental ones at different temperatures, finding an overall good agreement. The comparison of simulated and experimental macroscopic behaviour of plastic deformation and rupture also showed the possibility to indirectly estimate micro- and mesoscale material parameters. Both heat conduction and elastic behaviour of VPS-W have been extensively investigated. New capabilities of the Image-Based FEM could be shown: decomposition of the heat transfer reduction as due to the individual morphological phases and back-fitting of the reduced stiffness at interlamellar boundaries. The

  7. Incinerators and health. guide for the behavior to have during a local demand of sanitary investigations around a domestic refuse incinerator; Incinerateurs et sante. Guide pour la conduite a tenir lors d'une demande locale d'investigations sanitaires autour d'un incinerateur d'ordures menageres

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-12-15

    11,4 million tons of municipal solid and assimilated waste were incinerated in France in 2000. The 123 incinerators compliant with the Order in Council of January 25, 1991 have undergone significant modifications in the last years, and the incineration techniques used are of great concern to the public. The backfitting to new regulations and the many research works have answered some of the rightful questions of the population on health risks caused by waste incineration. However, many doubts remain and there has been many requests by the local population for epidemiological investigations to be conducted on this issue. The objectives of this document, requested by the Health General Directorate and presented as 'actions to be taken', are to inform the decentralized services of the government and regional epidemiology units of the health problems caused by waste incineration facilities and to help them grasp on a local level the situation met around these facilities. Therefore, this paper provides some scientific arguments to justify the need (or not) for setting up some specific studies as part of an informed public health management. This document is divided in three parts. The first part describes the actions to be taken at the local level. The methodological framework is based on: i) an analysis of the local situation; ii) finding a new definition in terms of public health to the one or more questions raised, and the usefulness to set up one or more health investigations; iii) the relevance of a specific type of study which would allow to answer these questions; and iv) the feasibility of this type of study. The second part briefly describes the various types of health studies and their use as a decision-making tool on waste-incineration facilities. These results stem mainly from the analysis of studies already put forward and carried out in past local situations. The third part points out what is currently found in today's literature on

  8. Atomic energy law after the opt-out. Alive and fascinating. Report about the 14th German atomic energy law symposium 2012

    International Nuclear Information System (INIS)

    Leidinger, Tobias

    2013-01-01

    Atomic energy law remains a living, fascinating subject matter. Nearly 200 participants were convinced of this impression at the 14 th German Atomic Energy Law Symposium held in Berlin on November 19-20, 2012. Under the scientific chairmanship of Professor Dr. Martin Burgi, Ludwig Maximilian University of Munich, the German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU), after an interruption of 5 years, again organized a scientific conference about practice-related topics of atomic energy and radiation protection law. Atomic energy law once again proved to be a reference area for sophisticated issues of constitutional law and administrative law above and beyond its technical confines. The agenda of the 14 th German Atomic Energy Law Symposium featured a broad spectrum of topics ranging from backfitting of nuclear power plants to European atomic energy and radiation protection law, to challenges facing national legal systems in the execution of atomic energy law, to legal issues connected with decommissioning and waste management, and on to the topical subject of finding a repository site. The 14 th German Atomic Energy Law Symposium, on the whole, again demonstrated that an open discourse between science and practice is able to furnish important contributions to the implementation of laws in a balanced way rooted in practice. Especially the contributions dealing with the independence of public authorities and their organization, the doctrine of the reservation of functions of the executive branch, and planning by laws contain additional provisions able to influence the continued development of administrative law also above and beyond atomic energy law. The BMU also referred to a decision just heard from Brussels to the effect that a new European Safety Directive would be published as early as in 2013. As a consequence of the nuclear stress tests conducted EU-wide, the Directive is to lay down provisions about transparency

  9. Including severe accidents in the design basis of nuclear power plants: An organizational factors perspective after the Fukushima accident

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Frutuoso e Melo, P.F.

    2015-01-01

    Highlights: • The Fukushima accident was man-made and not caused by natural phenomena. • Vulnerabilities were known by regulator and licensee but measures were not taken. • There was lack of independence and transparency of the regulatory body. • Laws and regulations have not been updated to international standards. • Organizational failures have played an important role in the Fukushima accident. - Abstract: The Fukushima accident was clearly an accident made by humans and not caused by natural phenomena as was initially thought. Vulnerabilities were known by both regulators and operator but they postponed measures. The emergency plan was not effective in protecting the public, because the involved parties were not sufficiently prepared to make the right decisions. The shortcomings and faults mentioned above resulted from the lack of independence and transparency of the regulatory body. Even laws and regulations, and technical standards, have not been upgraded to international standards. Regulators have not defined requirements and left for the operator to decide what would be more appropriate. In this aspect, there was clearly a lack of independence between these bodies and operator’s lobby power. The above situation raised the question of urgent updating of institutions, in particular those responsible for nuclear safety. The above evidences show that several nuclear safety principles were not followed. This paper intends to highlight some existing safety criteria that were developed from the operational experience of the severe accidents that occurred at TMI and Chernobyl that should be incorporated in the design of new nuclear power plants and to provide appropriate design changes (backfittings) for reactors that belong to the previous generation prior to the occurrence of these accidents, through the study of design vulnerabilities. Furthermore, the main criteria that define an effective regulatory agency are also discussed. Although these

  10. Aging and lifetime management - A plant-wide concept and examples for realization

    International Nuclear Information System (INIS)

    Erve, M.

    1998-01-01

    planning of maintenance and backfitting activities; the reduction of maintenance costs. Moreover, many investments can be coupled with an improvement in efficiency, uprating, or a combination of these. The concept works on four levels of different amounts of service integration: parts of components, components, systems, or whole plants. It has been applied so far to individual components and systems in Siemens/KWU plants and in plants of other system suppliers. Examples are presented in the paper. (author)

  11. Methods used to seismically upgrade. The safety related components of Belgian plants

    International Nuclear Information System (INIS)

    Lafaille, J.P.

    1993-01-01

    Belgian nuclear power amounts to about 6,000 MW, generated by seven plants that started operation as early as 1967. The latest plant started in 1985. Some of these plants were designed with no seismic requirements whatsoever. Even for those that had seismic requirements at the design stage, seismic demand was raised after design had been frozen (late during construction or at the 10 years revision). As a consequence all the plants had to undergo, to a variable extent, a seismic reevaluation and/or backfitting. Civil structures were concerned as well as electro-mechanical equipment and piping systems. The present paper deals with the mechanical aspect of the problem (equipment and piping). In order to minimize hardware modifications, advanced analytical techniques were used throughout the process, starting with the elaboration of a site specific spectrum, and using a full soil-structure interaction in order to get as 'realistic' as possible floor response spectra. In some instances, non linear elasto-plastic time history analysis was performed on piping-systems in order to qualify them without hardware modifications. In other cases a 'Load Coefficient Method' was used. Sometimes stresses or displacements taken from the original stress reports and scaled by comparison of applicable spectra, allowed to assess the seismic validity of the system under investigation. Seismic acceptability of installed active equipment is more difficult to demonstrate, as this is usually done by testing. This problem is a generic issue in the US, identified under the label USI-A-46 (Unresolved Safety Issue). It is treated by. a group of Utilities (SQUG = Seismic Qualification Utilities Group). The Belgian Utility is member of that group since 1985. The application of this program is starting in the US. SQUG methodology has been applied to three Belgian plants starting in 1988 and is now completed. The required fixes are being implemented. Experience gained in the process has been applied

  12. Hardware replacements and software tools for digital control computers

    International Nuclear Information System (INIS)

    Walker, R.A.P.; Wang, B-C.; Fung, J.

    1996-01-01

    Technological obsolescence is an on-going challenge for all computer use. By design, and to some extent good fortune, AECL has had a good track record with respect to the march of obsolescence in CANDU digital control computer technology. Recognizing obsolescence as a fact of life, AECL has undertaken a program of supporting the digital control technology of existing CANDU plants. Other AECL groups are developing complete replacement systems for the digital control computers, and more advanced systems for the digital control computers of the future CANDU reactors. This paper presents the results of the efforts of AECL's DCC service support group to replace obsolete digital control computer and related components and to provide friendlier software technology related to the maintenance and use of digital control computers in CANDU. These efforts are expected to extend the current lifespan of existing digital control computers through their mandated life. This group applied two simple rules; the product, whether new or replacement should have a generic basis, and the products should be applicable to both existing CANDU plants and to 'repeat' plant designs built using current design guidelines. While some exceptions do apply, the rules have been met. The generic requirement dictates that the product should not be dependent on any brand technology, and should back-fit to and interface with any such technology which remains in the control design. The application requirement dictates that the product should have universal use and be user friendly to the greatest extent possible. Furthermore, both requirements were designed to anticipate user involvement, modifications and alternate user defined applications. The replacements for hardware components such as paper tape reader/punch, moving arm disk, contact scanner and Ramtek are discussed. The development of these hardware replacements coincide with the development of a gateway system for selected CANDU digital control

  13. ACR-1000: Operator - based development

    International Nuclear Information System (INIS)

    Shalaby, B.; Alizadeh, A.

    2007-01-01

    Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU * reactors to establish Generation III + Advanced CANDU Reactor T M (ACR T M) technology. The ACR-1000 T M nuclear power plant is an evolutionary product, starting with the strong base of CANDU reactor technology, coupled with thoroughly-demonstrated innovative features to enhance economics, safety, operability and maintainability. The ACR-1000 benefits from AECL's continuous-improvement approach to design, that enabled the traditional CANDU 6 product to compile an exceptional track record of on-time, on budget product delivery, and also reliable, high capacity-factor operation. The ACR-1000 engineering program has completed the basic plant design and has entered detailed pre-project engineering and formal safety analysis to prepare the preliminary (non-project-specific) safety case. The engineering program is strongly operator-based, and encompasses much more than traditional pre-project design elements. A team of utility-experienced operations and maintenance experts is embedded in the engineering team, to ensure that all design decisions, at the system and the component level, are taken with the owner-operator interest in mind. The design program emphasizes formal review of operating feedback, along with extensive operator participation in program management and execution. Design attention is paid to layout and access of equipment, to component and material selection, and to ensuring maximum ability for on-line maintenance. This enables the ACR-1000 to offer a three-year interval between scheduled maintenance outages, with a standard 21-day outage duration. SMART CANDU T M technology allows on-line monitoring and diagnostics to further enhance plant operation. Modules of the Advanced CANDU SMART technologies are already being back-fitted to current CANDU plants. As well as reviewing the ACR-1000 design features and their supporting background, the paper describes the status of

  14. Implementation of the obligations of the Convention on Nuclear Safety CNS - Switzerland’s seventh national report to the Convention on Nuclear Safety

    International Nuclear Information System (INIS)

    2016-07-01

    In the aftermath of the Fukushima Daiichi accident in 2011, the Swiss government decided to phase out nuclear energy. Existing plants will continue to operate as long as they are considered safe by the Swiss Federal Nuclear Safety Inspectorate (ENSI) and as long as they fulfil all legal and regulatory requirements in this respect. In Switzerland, on-going activities regarding safety assessment of the different stages in the lifetime of nuclear installations consist of periodic assessments and assessments of long-term operation for existing Swiss nuclear power plants (NPPs). Assessments of long-term operation have been performed for two Swiss NPPs (Beznau and Muehleberg) which have been in commercial operation for over 40 years. A detailed examination demonstrated that the conditions for taking a NPP out of service have not yet been reached and will not be reached by these two plants within the next 10 years. Nevertheless, it is mandatory to continue with the scheduled ageing management, maintenance and backfitting activities. In late 2013, BKW Energy Ltd announced that Muehleberg NPP will be decommissioned at the end of 2019. The plant will shut down on December 20 th , 2019.The single 373 MWe boiling water reactor began operating in 1972. It will be the first Swiss nuclear power plant to be decommissioned. The preparatory work for decommissioning is well under way. In April 2015, a follow-up mission was conducted by the Integrated Regulatory Review Service in Switzerland. The Swiss government should give ENSI the ability to issue legally binding technical safety requirements and license conditions concerning nuclear safety, nuclear security and radiation safety. A follow-up mission by the Operational Safety Review Team on the Muehleberg NPP was completed in June 2014. Switzerland participated in the European Stress Test and its follow-up activities. During 2014, the necessary measures to achieve continuous improvement in the supervisory culture were defined. The

  15. Atomic energy law after the opt-out. Alive and fascinating. Report about the 14{sup th} German atomic energy law symposium 2012; Atomrecht nach dem Ausstieg. Lebendig und spannend. Tagungsbericht 14. Deutsches Atomrechtssymposium 2012

    Energy Technology Data Exchange (ETDEWEB)

    Leidinger, Tobias [Gleiss Lutz Rechtsanwaelte, Duesseldorf (Germany)

    2013-01-15

    Atomic energy law remains a living, fascinating subject matter. Nearly 200 participants were convinced of this impression at the 14{sup th} German Atomic Energy Law Symposium held in Berlin on November 19-20, 2012. Under the scientific chairmanship of Professor Dr. Martin Burgi, Ludwig Maximilian University of Munich, the German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU), after an interruption of 5 years, again organized a scientific conference about practice-related topics of atomic energy and radiation protection law. Atomic energy law once again proved to be a reference area for sophisticated issues of constitutional law and administrative law above and beyond its technical confines. The agenda of the 14{sup th} German Atomic Energy Law Symposium featured a broad spectrum of topics ranging from backfitting of nuclear power plants to European atomic energy and radiation protection law, to challenges facing national legal systems in the execution of atomic energy law, to legal issues connected with decommissioning and waste management, and on to the topical subject of finding a repository site. The 14{sup th} German Atomic Energy Law Symposium, on the whole, again demonstrated that an open discourse between science and practice is able to furnish important contributions to the implementation of laws in a balanced way rooted in practice. Especially the contributions dealing with the independence of public authorities and their organization, the doctrine of the reservation of functions of the executive branch, and planning by laws contain additional provisions able to influence the continued development of administrative law also above and beyond atomic energy law. The BMU also referred to a decision just heard from Brussels to the effect that a new European Safety Directive would be published as early as in 2013. As a consequence of the nuclear stress tests conducted EU-wide, the Directive is to lay down provisions about

  16. Studies on the effectiveness of measures to maintain the integrity of pressurized components in German nuclear power plants. Final report; Untersuchungen zur Wirksamkeit von Massnahmen zur Sicherstellung der Integritaet druckfuehrender Komponenten in deutschen Kernkraftwerken. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Elmas, M.; Jendrich, U.; Michel, F.; Reck, H.; Schimpfke, T.; Walter, M.; Wenke, R.

    2013-03-15

    The overall objective of the project was to investigate the effectiveness of measures to maintain the as-built quality of the pressure-retaining components in German nuclear power plants. In particular, investigations were performed on the application of the break preclusion concept, existing monitoring systems and the significance of the pressure test as part of the inspection concept. Moreover, the KompInt knowledge base has been updated. Break preclusion for pipes was applied in all German plants already during planning or after commissioning to a varying extent. The basic features of the required assessments were considered in the German nuclear regulations for the first time by inclusion in the safety requirements for nuclear power plants of 2012. The requirements for assessments, differing in their degree of detail, in the interpretations of these safety requirements and in the safety standard KTA 3206 are still in the draft stage. For the first time, the vessels as well as housings of valves and pumps are also included in the concept. Through the use of advanced monitoring systems it was possible in German plants at an early stage to establish modes of operation that minimise the load on components, to carry out appropriate technical backfitting measures, and to identify damages. In plant areas where local water chemistry parameters may result that deviate from the specification, the effectiveness of water chemistry monitoring is limited. In this case, other operational measures must be taken. The results of the simulations performed with the help of the GRS-developed PROST computer code to determine the significance of pressure tests lead - in accordance with the results of operating experience evaluation - to the conclusion that pressure tests carried out within the pressure-retaining boundary contribute to safeguarding the integrity. The user-friendliness of the KompInt knowledge base has been increased by changing over to a new hardware, a software

  17. Debris impact on emergency coolant recirculation - summary and conclusions

    International Nuclear Information System (INIS)

    Jain, Bhagwat; Hsia, Anthony; Armand, Yves; Mattei, Jean-Marie; Hyvaerinen, Juhani; Maqua, Michael; Puetter, Bernhard; Sandervaag, Oddbjoern; Vandewalle, Andre; Tombuyses, Beatrice; Pyy, Pekka; Royen, Jacques

    2004-01-01

    On 28 July 1992, a steam line safety relief valve inadvertently opened in the Barsebaeck-2 nuclear power plant in Sweden. The steam jet stripped fibrous insulation from adjacent piping system. Part of that insulation debris was transported to the wet-well pool and clogged the intake strainers for the drywell spray system after about one hour. Although the incident in itself was not very serious, it revealed a weakness in the defense-in-depth concept which under other circumstances could have led to the emergency core cooling system (ECCS) failing to provide recirculation water to the core. The Barsebaeck incident spurred immediate action on the part of regulators and utilities alike in several OECD countries. Research and development efforts of varying degrees of intensity were launched in many countries and in several cases resulted in findings that earlier strainer clogging data were incorrect because essential parameters and physical phenomena had not been recognized previously. Such efforts resulted in substantial back-fittings being carried out for BWRs and some PWRs in several OECD countries. An international workshop organised in Stockholm in 1994 under the auspices of CSNI revealed a rather confusing picture of the available knowledge base, examples of conflicting information and a wide range of interpretation of guidance for assessing BWR strainers and PWR sump screen performance contained in US NRC Regulatory Guide 1.82. An International Working Group was set up by the CSNI to establish an internationally agreed-upon knowledge base for assessing the reliability of ECC water recirculation systems. An initiative was taken by the CSNI in 1998 to revisit the subject. The general objective was to make an update of the knowledge base for strainer clogging, to review the latest phenomena for PWRs and to provide a survey of actions taken in member countries. New information contained in NUREG/CR-6771 indicated that the core damage frequency could increase by one

  18. Proceedings (slides) of the OECD/NEA Workshop on Innovations in Water-cooled Reactor Technologies

    International Nuclear Information System (INIS)

    Spiler, Joze; Kim, Sang-Baik; ); Feron, Fabien; Jaervinen, Marja-Leena; Husse, Julien; ); Ferraro, Giovanni; Bertels, Frank; Denk, Wolfgang; Tuomisto, Harri; Golay, Michael; Buongiorno, J.; Todreas, N.; Adams, E.; Briccetti, A.; Jurewicz, J.; Kindfuller, V.; Srinivasan, G.; Strother, M.; Minelli, P.; Fasil, E.; Zhang, J.; Genzman, G.; Epinois, Bertrand de l'; Kim, Shin Whan; Laaksonen, Jukka; Maltsev, Mikhail; Yu, CHongxing; Powell, David; Gorgemans, Julie; Hopwood, Jerry; Bylov, Igor; Bakhmetyev, Alexander M.; Lepekhin, Andrey N.; Fadeev, Yuriy P.; Bruna, Giovanni; Gulliford, Jim; ); Ham-Su, Rosaura; Thevenot, Caroline; GAUTIER, Guy-Marie; MARSAULT, Philippe; PIGNATEL, Jean-Francois; White, Andrew; )

    2015-02-01

    New technologies and solutions have been developed over more than thirty years to improve the safety, performance and economics of nuclear power plants. Particular efforts were made in designing systems to prevent or mitigate nuclear accidents and, greatly limit or even avoid any offsite release of radioactivity. Reactor designs developed in the 1980's and later are often referred to as Generation III (Gen III) reactors. They offer enhanced safety compared to earlier Generation II (Gen II) designs, as well as improved performance and economics. Examples of Gen III safety design features include solutions for corium localisation, advanced containment structures, improved emergency core-cooling systems, filtered venting systems, hydrogen risk management solutions, etc. Some of these solutions have also been back-fitted or partially adapted to existing reactors, based on recommendations from regulators or modernisation efforts by the utilities operating these reactors, to bring their level of safety to levels approaching those of the more modern designs. Other innovations found in the latest water-cooled reactor designs include the use of passive safety systems, and often associated with those, a simplification in the design of the reactor. Gen III reactors also feature better economics, for example increased design lifetime up to 60 years, ability to use 100% MOX fuel and operate with higher flexibility, higher thermal efficiencies and reduced staff requirements. Modularity is often quoted as a feature of some Gen III designs as a way of reducing the construction times and simplifying the decommissioning of the plant. The scope of the Workshop includes, inter alia: - Evolution of regulatory and design requirements for commercial water-cooled reactors; - Innovations in water-cooled reactor technologies that allowed significant improvement in the level of safety, with a discussion on advantages and challenges of active vs. passive safety systems; - Innovations under

  19. Paul Scherrer Institut Scientific Report 2002. Volume IV: Nuclear Energy and Safety

    International Nuclear Information System (INIS)

    Smith, B.; Gschwend, B.

    2003-03-01

    the FIT Board. The planning document is currently in circulation for comment. PSI has updated its 'Guiding Principles' to include its own research capabilities, in particular in regard to sustainable energy technologies. At the end of an intensive 'bottom-up approach', an R+D planning document for the year 2003 has been issued for the first time by NES; this exercise should be formalised in a more compact form in the future. Backfitting of the Hot Lab has been completed in 4 years, at a cost of 14 MCHF (compared with the initial estimates of 3 years and 9.8 MCHF); the NPP operators have agreed to contribute to the Hot Lab infrastructure costs to the extent of 3 MCHF per annum. With regard to education and training, in view of the necessity to maintain a continuous education programme in nuclear technology at university level, the Swiss utilities have conditionally agreed, following the retirement of Prof. Yadigaroglu, to fund an ETHZ Chair in Nuclear (Systems) Engineering. The appointee will become a central part of the new Master of Nuclear Engineering programme, to be offered by ETHZ and EPFL, and will be given the opportunity by NES to perform large-scale research at PSI. The new Masters programme will become an integral part of the emerging European Network on Nuclear Energy (ENEN). At the technical teaching level, PSI's own Technical School for reactor operators has, after some years of reduced participation, again registered an increase in the number of students: in 2002, courses had to be offered on an annual basis, instead of every 2 years, in response to the increased demand. A publication list for 2002 is also provided. (author)

  20. International Nuclear Safety Experts Conclude IAEA Peer Review of Swiss Regulatory Framework

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: A team of international nuclear safety experts today completed a two-week International Atomic Energy Agency (IAEA) review of the regulatory framework for nuclear safety in Switzerland. The Integrated Regulatory Review Service (IRRS) mission noted good practices in the Swiss system and also made recommendations for the nation's nuclear regulatory authority, the Swiss Federal Nuclear Safety Inspectorate (ENSI). ''Our team developed a good impression of the independent Swiss regulator - ENSI - and the team considered that ENSI deserves particular credit for its actions to improve Swiss safety capability following this year's nuclear accident in Japan,'' said IRRS Team Leader Jean-Christophe Niel of France. The mission's scope covered the Swiss nuclear regulatory framework for all types of nuclear-related activities regulated by ENSI. The mission was conducted from 20 November to 2 December, mainly at ENSI headquarters in Brugg. The team held extensive discussions with ENSI staff and visited many Swiss nuclear facilities. IRRS missions are peer reviews, not inspections or audits, and are conducted at the request of host nations. For the Swiss review, the IAEA assembled a team of 19 international experts from 14 countries. The experts came from Belgium, Brazil, the Czech Republic, Finland, France, Germany, Italy, the Republic of Korea, Norway, Russia, Slovakia, Sweden, the United Kingdom, and the United States. ''The findings of the IRRS mission will help us to further improve our work. That is part of our safety culture,'' said ENSI Director General Hans Wanner. ''As Switzerland argued at international nuclear safety meetings this year for a strengthening of the international monitoring of nuclear power, we will take action to fulfil the recommendations.'' The IRRS team highlighted several good practices of the Swiss regulatory system, including the following: ENSI requires Swiss nuclear operators to back-fit their facilities by continuously upgrading

  1. Paul Scherrer Institut Scientific Report 2002. Volume IV: Nuclear Energy and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Smith, B.; Gschwend, B. (eds.)

    2003-03-01

    contributions have been accepted by the FIT Board. The planning document is currently in circulation for comment. PSI has updated its 'Guiding Principles' to include its own research capabilities, in particular in regard to sustainable energy technologies. At the end of an intensive 'bottom-up approach', an R+D planning document for the year 2003 has been issued for the first time by NES; this exercise should be formalised in a more compact form in the future. Backfitting of the Hot Lab has been completed in 4 years, at a cost of 14 MCHF (compared with the initial estimates of 3 years and 9.8 MCHF); the NPP operators have agreed to contribute to the Hot Lab infrastructure costs to the extent of 3 MCHF per annum. With regard to education and training, in view of the necessity to maintain a continuous education programme in nuclear technology at university level, the Swiss utilities have conditionally agreed, following the retirement of Prof. Yadigaroglu, to fund an ETHZ Chair in Nuclear (Systems) Engineering. The appointee will become a central part of the new Master of Nuclear Engineering programme, to be offered by ETHZ and EPFL, and will be given the opportunity by NES to perform large-scale research at PSI. The new Masters programme will become an integral part of the emerging European Network on Nuclear Energy (ENEN). At the technical teaching level, PSI's own Technical School for reactor operators has, after some years of reduced participation, again registered an increase in the number of students: in 2002, courses had to be offered on an annual basis, instead of every 2 years, in response to the increased demand. A publication list for 2002 is also provided. (author)

  2. Aging evaluation methodology of periodic safety review in Korea

    International Nuclear Information System (INIS)

    Park, Heung-Bae; Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In Korea plant lifetime management (PLIM) study for Kori Unit 1 has been performed since 1993. Meanwhile, periodic safety review (PSR) for all operating nuclear power plants (NPPs) has been started with Kori Unit 1 since 2000 per IAEA recommendation. The evaluation period is 10 years, and safety (evaluation) factors are 11 per IAEA guidelines as represented in table 1. The relationship between PSR factors and PLIM is also represented. Among these factors evaluation of 'management of aging' is one of the most important and difficult factor. This factor is related to 'actual condition of the NPP', 'use of experience from other nuclear NPPs and of research findings', and 'management of aging'. The object of 'management of aging' is to obtain plant safety through identifying actual condition of system, structure and components (SSCs) and evaluating aging phenomena and residual life of SSCs using operating experience and research findings. The paper describes the scope and procedure of valuation of 'management of aging', such as, screening criteria of SSCs, Code and Standards, evaluation of SSCs and safety issues as represented. Evaluating SSCs are determined using final safety analysis report (FSAR) and power unit maintenance system for Nuclear Ver. III (PUMAS/N-III). The screening criteria of SSCs are safety-related items (quality class Q), safety-impact items (quality class T), backfitting rule items (fire protection (10CFR50.48), environmental qualification (10CFR50.49), pressurized thermal shock (10CFR50.61), anticipated transient without scram (10CFR50.62), and station blackout (10CFR50.63)) and regulating authority requiring items[1∼3]. The purpose of review of Code and Standards is identifying actual condition of the NPP and evaluating aging management using effective Code and Standards corresponding to reactor facilities. Code and Standards is composed of regulating laws, FSAR items, administrative actions, regulating actions, agreement items, and other

  3. Experience and trends at the Belgonucleaire plant

    International Nuclear Information System (INIS)

    Deramaix, P.; Eeckhout, F.; Pay, A.; Pelckmans, E.

    2000-01-01

    after 6 irradiation cycles. No failures due to the MOX were noticed. Since the mid 1990's, the plant is being backfitted without interruption of the fabrication, to incorporate improvements resulting from accumulated experience to improve still further the flexibility of the plant while meeting the more challenging future requirements in particular in terms of radioprotection regulation (degraded plutonium isotopic composition, higher burnup design fuel assemblies, ICRP 60) as well as in terms of economics (recycling of the scrap, reduction of the fabrication generated waste). (author)

  4. EU-stress test: Swiss national action plan. Follow-up of peer review 2012 year-end status report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    The European Union (EU) stress test is part of the review process which Switzerland initiated immediately after the reactor accident in Japan. As a direct consequence of the accident, the Swiss Federal Nuclear Safety Inspectorate (ENSI) issued three formal orders in which the operators of the Swiss nuclear power plants (NPPs) were required to implement immediate measures and to conduct additional reassessments. The immediate measures comprised the establishment of an external emergency storage facility for the Swiss NPPs, including the necessary plant-specific connections, and back-fittings to provide external injection into the spent fuel pools. The additional reassessments focused on the design of the Swiss NPPs against earthquakes, external flooding and a combination thereof. Investigations on the coolant supply for the safety systems and the spent fuel pool cooling were also requested. ENSI carried out an analysis of the events at Fukushima providing detailed descriptions of the causes, consequences and radiological impacts of the accident. The reports analyse the contributory human and organisational factors, and specify lessons that can be derived from this information. ENSI instructed the Swiss operators to take part in the EU stress test. There was to be particular examination of the robustness of the NPPs in case of impacts beyond the design basis due to earthquakes, external flooding and extreme weather conditions, with consequential loss of power supply and heat sink, and the need for severe accident management actions. ENSI requested further clarification on plant specific issues and produced the National Report which was delivered to the EU Commission. A Country Peer Review Draft Report was drawn up for each country, including a list of issues (open points) for further follow-up by the review team. Eight new open points were identified to further improve the safety of the Swiss NPPs. These open points together with the issues identified in the analysis

  5. Financing long term liabilities (Germany)

    International Nuclear Information System (INIS)

    2003-01-01

    implementation of the measures will cover a period of 15 to 20 years depending on the site. The necessary expenses are carried by the Federal Government and estimated to amount to about EUR 6.5 billion. In addition the Federal Republic of Germany inherited 6 operating NPPs of soviet design from the former GDR. Comprehensive safety analyses after the German reunification arrived at the conclusion that they did not correspond to Western German safety standards. They had to be shut down in 1990. As the power industry was not prepared to carry the financial risks of backfitting and re-licensing the reactors, the Federal Republic of Germany took over the liabilities. The aim is to finish the decommissioning activities around the year 2012. The total costs for dismantling the plants and storing the resulting waste are estimated to amount to about EUR 3.1 billion

  6. Implementation of the obligations of the convention on nuclear safety. Fourth Swiss report in accordance with Article 5

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-15

    . Emergency drills are conducted at regular intervals. The international alerting system is also in a mature stage. The first generation of NPPs in Switzerland has been the subject of progressive back-fitting. The second generation of NPPs incorporated various safety and operating improvements in their initial design. All Swiss NPPs have undergone the safety review process required under the Convention and have incorporated the improvements identified in the respective safety review reports. The Swiss policy of continuous improvements to NPPs ensures a high level of safety. The legislation and regulatory framework for nuclear installations is well established. It provides the formal basis for the supervision and the continuous improvement of nuclear installations. The supervisory authority conducts inspections and technical discussions with the utilities to ensure that operators assume full responsibility for the safety of their installations. All NPPs have implemented programmes to improve their safety culture. Plant-specific full scope replica simulators are operating at all Swiss NPPs. The Inspectorate's organisation includes staff members dealing with human aspects, NPP organisation, and safety culture. Considerable attention is paid to human factor aspects of operator support systems, including procedures, guidelines and checklists. The review and assessment procedure includes an evaluation of the safety analysis report, safety-relevant systems, design-basis accident analyses, probabilistic safety analysis and reports on ageing surveillance programmes. An Ageing Surveillance Programme is in place for all NPPs in order to maintain safety margins and safety functions of structures, systems and components throughout the plant lifetime. Concerning the radiation protection, the supervisory and control methods currently applied by the inspectorate are in compliance with the Convention's requirement to keep radioactive doses to the public and the environment as low as

  7. Innovative probabilistic risk assessment applications: barrier impairments and fracture toughness. 2. Demolition Debris and Tornado Missile Hazard During Decommissioning

    International Nuclear Information System (INIS)

    Calhoun, David; Shepherd, Stephen

    2001-01-01

    impose on decommissioning projects. Unit 1 began operation in 1968. Because of the age of Unit 1's design and the low frequency of tornadoes in California, the original plant design did not provide any protection from tornado hazards. Tornado protection requirements were later imposed as a back-fit; however, the approved license change was based on a probabilistic risk assessment that defined Unit 1's tornado missile damage acceptance limit in terms of reactor core damage frequency. When several Unit 1 buildings have been demolished, construction will begin on an ISFSI for Unit 1's spent fuel. The ISFSI design incorporates tornado missile barrier features into the storage canister and transfer cask. These design provisions will alleviate any need to manage tornado missile hazards. Units 2 and 3 share a design basis for tornado missile protection that closely follows the U.S. Nuclear Regulatory Commission's Standard Review Plan (NUREG 0800), Revision 1. Critical components are identified that are required to be functional following design-basis tornadoes. Missile barriers protect most critical components; however, some critical components are allowed to be exposed to tornado missiles provided the aggregate annual probability of damage to all critical components is -7 per unit. According to the analysis that established this probability, it is directly proportional to the inventory of unrestrained objects within a missile pickup/transport area that includes the entire site. To determine the increased probability of damage due to demolition work, the quantity of loose debris was estimated for several discrete time intervals of the decommissioning process. This intermediate result showed that debris controls would be necessary to protect critical components in Units 2 and 3 during the demolition of Unit 1. Several different methods for controlling debris were evaluated for efficacy, feasibility, and cost-effectiveness. Unit 1 decommissioning work will increase the number of

  8. Implementation of the obligations of the convention on nuclear safety. Fifth Swiss report in accordance with Article 5

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-15

    Switzerland signed the Convention on Nuclear Safety (CNS). In accordance with Article 5 of CNS, Switzerland has submitted 4 country reports for Review Meetings of Contracting Parties. This 5{sup th} report by the Swiss Federal Nuclear Safety Inspectorate (ENSI) provides an update on compliance with CNS obligations. The report attempts to give appropriate consideration to issues that aroused particular interest at the 4{sup th} Review Meeting. It starts with general political information on Switzerland, a brief history of nuclear power and an overview of Swiss nuclear facilities. This is followed by a comprehensive overview of the status of nuclear safety in Switzerland (as of July 2010) which indicates how Switzerland complies with the key obligations of the Convention. ENSI updated a substantial proportion of its guidelines which are harmonised with the safety requirements of the Western European Nuclear Regulators Association (WENRA) based on IAEA Safety Standards. On 1{sup st} January 2009, ENSI became formally independent of the Swiss Federal Office of Energy. It is now a stand-alone organisation controlled by its own management board. Switzerland recently started a process to select a site for the disposal of radioactive waste in deep geological formations. The first generation of NPPs in Switzerland has been the subject of progressive back-fitting. The second generation of NPPs incorporated various safety and operating improvements in their initial design. All Swiss NPPs have undergone the safety review process required under the Convention and have incorporated the improvements identified in the respective safety review reports. The Swiss policy of continuous improvements to NPPs ensures a high level of safety. The legislation and regulatory framework for nuclear installations is well established. It provides the formal basis for the supervision and the continuous improvement of nuclear installations. The Nuclear Energy Act and its ordinance came into force

  9. Critical review of the national action plans (NAcP) of the EU stress tests on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Oda; Lorenz, Patricia

    2013-04-15

    preparation and again later backfittings would start. In some cases this is officially scheduled to

  10. Implementation of the obligations of the convention on nuclear safety. Fourth Swiss report in accordance with Article 5

    International Nuclear Information System (INIS)

    2007-07-01

    conducted at regular intervals. The international alerting system is also in a mature stage. The first generation of NPPs in Switzerland has been the subject of progressive back-fitting. The second generation of NPPs incorporated various safety and operating improvements in their initial design. All Swiss NPPs have undergone the safety review process required under the Convention and have incorporated the improvements identified in the respective safety review reports. The Swiss policy of continuous improvements to NPPs ensures a high level of safety. The legislation and regulatory framework for nuclear installations is well established. It provides the formal basis for the supervision and the continuous improvement of nuclear installations. The supervisory authority conducts inspections and technical discussions with the utilities to ensure that operators assume full responsibility for the safety of their installations. All NPPs have implemented programmes to improve their safety culture. Plant-specific full scope replica simulators are operating at all Swiss NPPs. The Inspectorate's organisation includes staff members dealing with human aspects, NPP organisation, and safety culture. Considerable attention is paid to human factor aspects of operator support systems, including procedures, guidelines and checklists. The review and assessment procedure includes an evaluation of the safety analysis report, safety-relevant systems, design-basis accident analyses, probabilistic safety analysis and reports on ageing surveillance programmes. An Ageing Surveillance Programme is in place for all NPPs in order to maintain safety margins and safety functions of structures, systems and components throughout the plant lifetime. Concerning the radiation protection, the supervisory and control methods currently applied by the inspectorate are in compliance with the Convention's requirement to keep radioactive doses to the public and the environment as low as reasonably achievable and also

  11. Sweden's second national report under the Convention on nuclear safety. Swedish implementation of the obligations of the Convention

    International Nuclear Information System (INIS)

    2001-01-01

    organisations, such as downsizing, outsourcing and merging, need to be followed closely, by the licensees as well as by the regulatory bodies, and methods need to be further developed to assess the safety consequences of such changes. The ongoing dialogue between the licensees and the regulator regarding development of safety in existing reactors needs to be concluded, in order to define reasonable requirements for back-fitting during the remaining operating time. The general concern expressed in the first report to the Convention about the shortage of qualified, university trained engineers and researchers in specific nuclear fields, still remains in the longer perspective. A plan is also under discussion to ensure financing of nuclear education and research at several universities for at least a three year period. Taking into account all these efforts, an action plan should be developed to ensure the necessary long-term nuclear competence in Sweden. At the first review meeting in April 1999, Sweden accepted to report on the following issues in particular, in its next report: 1. measures to upgrade the older reactors and how these comply with the safety regulations, 2. measures within the industry and the regulatory bodies to improve the safety culture, 3. monitoring of the effects, if any, on safety as a consequence of deregulation of the electricity market, 4. experience gained from the new safety regulations, especially with regard to the higher requirements placed on the licensees own control over safety. These reports do not indicate any concerns as to the Swedish compliance with the obligations under the Convention

  12. Swiss regulatory use of databanks for nuclear power plant life management, surveillance and safety analyses

    International Nuclear Information System (INIS)

    Tipping, Ph.; Beutler, R.; Schoen, G.; Noeggerath, J.

    2002-01-01

    achieved. The Inspectorate processes these data under the auspices of its own specialist group, and a final decision as to the root cause and the safety importance is made. In this way, any differences in interpretation of importance and safety impact of events between the Inspectorate's own assessment and that of the NPP operators can be analysed, discussed and put into the correct context and perspective. Generally, the reportable event assessments and proposed mitigation or other actions of the operators have been found to be acceptable to the Inspectorate, but, in some cases, differences between the interpretations of the regulator and operator have become apparent. The Inspectorate has, over the years, collected data concerned with all aspects of safety, backfitting and modifications in the Swiss and also other NPPs. The main DBs of the Inspectorate are: 1) Reportable Events DB, 2) Probabilistic Safety Analysis (PSA) DB and 3) Damage and Degradation of SSCs DB. The Inspectorate's reportable events DB has been conceived to incorporate a classification of SSCs and failure types according to the IAEA/NEA incident reporting system (IRS). All the DBs enable the user to obtain condensed reports of the incidents, materials and systems or components involved, the assessments of the NPP operators and the finally binding, salient points and lessons-learned summaries with recommendations or requirements to the NPP operator, from the Inspectorate. All of the DBs are updated regularly since they are living documents. The DBs are so conceived that the Swiss NPPs (Muehleberg/G.E.BWR; Beznau 1 and 2 /Westinghouse PWRs; Goesgen/KWU PWR and Leibstadt/G.E. BWR) can be individually analysed and, where applicable, comparisons undertaken. The Inspectorate's DBs have proven to be informative and practical tools to register, monitor and register information on all events concerned with the operation of NPPs. An overview of the structures of the individual DBs is provided. Focus is made on

  13. Generation IV nuclear energy systems: road map and concepts. 2. Generation II Measurement Systems for Generation IV Nuclear Power Plants

    International Nuclear Information System (INIS)

    Miller, Don W.

    2001-01-01

    , humidity, smoke, and high temperature). Reference 4 describes the use of a Fabry-Perot fiber-optic temperature sensor that was selected for performance evaluation and for potential application in nuclear power plants because of its unique interferometric mechanism and data processing technique and its commercial availability. In the past several years, the use of acoustic methods, either transmission timing or correlation methods, have been developed to the point that they are being introduced as a back-fit in operating plants. The advantage these methods offer is increased accuracy, which translates into increased reactor power. A new method for local measurement of reactor power is being developed at Ohio State. This power sensor concept is based on maintaining a constant temperature in a small mass of actual reactor fuel or fuel analogue by adding heat through resistive dissipation of input electrical energy. Sensors of this type can provide a direct measurement of the nuclear energy deposition rather than neutron flux. Holcomb at Oak Ridge National Laboratory is proposing to develop a combined optical-based neutron flux/temperature sensor for in-core measurements in high-temperature gas reactors. The current status of I and C systems in nuclear power plants was reviewed, and it was concluded that the fundamental measuring systems had not changed substantially since the early nuclear plants. New methods and advanced measuring systems were discussed. Advanced systems of the type discussed should be considered in the design of next-generation I and C systems. However, they should be considered along with the sensors and systems currently being used, which have served their functions very well for the past 40 yr. (authors)