WorldWideScience

Sample records for backfitting

  1. Backfitting guidelines

    International Nuclear Information System (INIS)

    The backfitting process is the process by which the US Nuclear Regulatory Commission (NRC) decides whether to issue new or revised requirements or staff positions to licensees of nuclear power reactor facilities. Backfitting is expected to occur and is an inherent part of the regulatory process. However, it is to be done only after formal, systematic review to ensure that changes are properly justified and suitably defined. Requirements for proper justification of backfits and information requests are provided by two NRC rules, Title 10 of the Code of Federal Regulations, Sections 50.109 and 50.54(f). Three types of backfits are recognized. Cost-justified substantial safety improvements require backfit analyses and findings of substantial safety improvement and justified costs. Two types of exceptions, compliance exceptions and adequate protection exceptions, do not require findings of substantial safety improvements and costs are not considered. However, they are still backfits and they require documented evaluations to support use of the exceptions. Information requests (as opposed to backfits) require an analysis of the burden to be imposed to ensure that they are justified in view of the potential safety significance of the information requested. NRC procedures on backfitting include the Charter of the Committee to Review Generic Requirements for generic communications and NRC Manual Chapter 0514 and individual office procedures for plant-specific communications. Considerable guidance has been developed, control mechanisms are in place, and training has been provided to NRC and industry personnel

  2. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  3. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    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 U3Si2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

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

    International Nuclear Information System (INIS)

    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

  5. The Alara principle in backfitting Borssele

    International Nuclear Information System (INIS)

    An extensive backfitting program, the Modifications Project, was carried out at the Borssele Nuclear Power Station. It involved sixteen modifications to technical systems. The scope of activities, and the dose rates encountered in places where work was to be performed, made it obvious from the outset that a high collective dose had to be anticipated. As a consequence, radiation protection within the project was organized in such a way that applicable radiation protection principles were applied in all phases of the project. From the point of view of radiation protection, the Modifications Project had to be subdivided into three phases, i.e., a conceptual design phase in which mainly the justification principle was applied; the engineering phase in which the Alara principle was employed; the execution phase in which management of the (internal) dose limits had to be observed in addition to the Alara principle. Throughout all project phases, radiation protection considerations and results were documented in so-called Alara reports and radiation protection checklists. As a result of the strictest possible observance of radiation protection principles in all phases of the project, a collective dose of 2505 mSv was achieved, which stands for a reduction by a factor of 4 compared to the very first estimate. In view of the scope and complex nature of the activities involved, and the radiation levels in the Borssele Nuclear Power Station, this is an excellent result. (orig.)

  6. Backfitting of I and C systems important to safety

    International Nuclear Information System (INIS)

    Replacing existing I and C systems requires special planning. This aims at significantly reducing the work that is to be carried out when the system is installed in the plant. With the new digital I and C TELEPERM XS system functionality can be extensively validated beforehand in various simulation environments and tests can be transferred to the test bay. This renders possible a backfit to system in a time which could not be attained with conventional technology. The individual steps of a successful integration and installation are presented for the example of the backfitting in the nuclear power plant Neckarwestheim unit 1. Together customer and supplier elaborated a schedule in close cooperation thus making possible a backfitting without having to extend the planned revision period. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  9. Backfitting in the Beznau and Muehleberg nuclear power plants

    International Nuclear Information System (INIS)

    Backfitting work is curried out at the Beznau and Muehleberg nuclear power plants in order to meet the legal requirement of up-to-date technology. Among this work, there is the large project of installing additional, automatic after-heat removal systems which work reliably even in case of outside influences. (orig.)

  10. Backfit of a redesigned radwaste processing and solidification system

    International Nuclear Information System (INIS)

    The original design of the radwaste system for the Enrico Fermi Atomic Power Plant, Unit 2, had been overtaken by recent developments in processing and volume reduction/solidification technology as well as heightened awareness of ALARA concepts and experience with similar systems in operating plants. This paper describes the design and backfit of current technology liquid and solid radwaste processing systems, including an asphalt-based volume reduction and solidification system

  11. NPP advanced pipework technologies: recent backfitting projects and computational analyses

    International Nuclear Information System (INIS)

    Some recent international NPP projects involving advanced engineering services for existing installations, such as replacement of valves, containment penetrations and pipework as well as design and installations of pipe supports and dynamic restraints, are summarized. Examples of thermomechanical analyses of operational phenomena performed as part of comprehensive plant design, licensing, and commissioning support activities are presented. Turnkey project management with system function warranty resulted in most effective use of all resources drawn together from several high-qualified subcontractors and international equipment manufacturers working under the supervision of an in-house team. Experience collected to date in backfitting various plants of different design and age provides a strong knowledge basis. It is available for evaluating any plant currently in operation or under construction in order to check the need for modifications and recommend the appropriate scheduling and level of effort. (authors)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  15. PSA based plant modifications and back-fits

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  17. Backfitting of nuclear power plants, as seen by the operator or manufacturer

    International Nuclear Information System (INIS)

    Backfitting is understood to mean a technically justified improvement of an existing plant or component, rather than a measure to adjust a plant to recent regulatory requirements. The Philippsburg-1 reactor is taken as an example to explain the entire backfitting process as seen by an operator, from the reasons leading to the first decision to the planning of type and extent of work, costs, and procedural planning. The activities described correspond to the work at the Philippsburg-1 reactor in the years 1980 and 1981. (DG)

  18. Assessment and inspection tasks of the Spanish regulatory body staff regarding I and C and related systems backfitting in old plants

    International Nuclear Information System (INIS)

    Based on our experience working on backfits of the two oldest plants in Spain, we believe that: (1) Reaching backfitting decisions is one of the most difficult tasks being performed by a regulatory body. (2) Any backfitting decision should be preceded by a thorough review of the safety importance of the situation the backfit is aimed to correct. (3) Backfitting decisions should be reached in an integrated way. A complete review of the plant should be performed to put each backfit in perspective. PRA may be a useful tool to achieve it. (4) Except when there is an immediate need of corrective actions backfitting schedules should be long enough to allow appropriate review by all involved parties

  19. Backfitting of independent residual heat-removal systems in West Germany and Switzerland

    International Nuclear Information System (INIS)

    The engineered solution to West Germany's requirements for coping with catastrophic events is an independent residual heat-removal (IRHR) system. The IRHR system must be physically separated from other plant equipment and have its own control room, heat sink, and power supply. Nuclear plants constructed since the mid-1970s are equipped with IRHR systems. Eight plants already in operation at that time required custom backfit designs, which were performed by Kraftwerk Union AG (KWU). The KWU systems are designed to cope with simultaneous loss of the main heat sink and power failure and in some cases a simultaneous loss-of-coolant accident. Backfitting generally takes 6 to 7 years with a cost of $50 to $90 million, including takes 6 to 7 years with a cost of $50 to $90 million, including 200,000 to 300,000 job-hours. Plant availability is not affected. 6 refs., 8 figs., 1 tab

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

    International Nuclear Information System (INIS)

    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

  1. Report on SARS backfit evaluation, Exxon Donor Solvent Plant, Baytown, Texas

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, A.F. Jr.

    1980-07-02

    This report provides information on observations, findings, and conclusions arising from a site visit to the Exxon Donor Solvent Plant, Baytown, Texas. That visit was to provide technical assistance and professional services to the DOE/ASFE/OPTA Project Officer regarding verification of his initial determination that this pilot plant is exempt from the SARS backfit requirement (DOE Order 5481.1). A secondary purpose was to obtain further information regarding the occupational safety and health plans and procedures at this new pilot plant facility. It is concluded that a well planned and implemented occupational safety and health program exists at the Exxon Donor Solvent Plant. Excellent manuals regarding general safety requirements and protection against carcinogens have been prepared and distributed. A Safe Operations Committee is in effect as is a Risk Management Committee. Adequate safety and industrial hygiene staff has been assigned and an excellent medical surveillance program has been established. Adequate compliance with environmental codes, standards, and regulations is being achieved. Although this plant is not subject to SARS because of the nature of the contract, adequate documentation exists in any case to exempt it from the SARS backfit requirement.

  2. The Belgian experience on the backfitting and safety upgrading of old operating nuclear power plants

    International Nuclear Information System (INIS)

    The paper describes the methodology for backfitting and safety upgrading during the reevaluation of the Belgian NPP's: first generation (Doel-1, Doel-2, Tihange-1) and second generation plants (Doel-3, Doel-4, Tihange-2 and Tihange-3). A list of essential safety subjects and topics is given. The experience has proved the feasibility of a safety upgrading of operating NPP without injury to its availability, the benefit of a close cooperation between owner, engineering company and safety authorities throughout the project. A global approach to solving numerous specific deficiencies along with the optimization of the investments regarding the safety improvement of the NPP is suggested. Further increase of the know-how will be achieved through the present Belgian programme along with similar activities abroad. (R.I.)

  3. Compensatiopn for loss as a result of revocation or withdrawal of license, or of post-licensing, imposition of backfits

    International Nuclear Information System (INIS)

    The paper reviews the current legal provisions, which under paragraph 18 of the Atomic Energy Act contain the principle of compensation to be paid in case of a revocation or withdrawal of a licence. Even in the current regime, post-licensing imposition of backfitting obligations is subject to indemunification only if the backfits represent an unreasonable burden to the plant operator. The proposed amendment of the AtG abandons this principle altogether. This is in agreement with constitutional law. But taking into account the interests of environmental law closely linked with the problem, a thorough revision of paragraph 18 AtG would have been the proper way of tackling problems involved with old nuclear installations. (HP)

  4. Does a reactor need a safety backfit? Case study on communicating decision and risk analysis information to managers

    International Nuclear Information System (INIS)

    An approach to communicating decision and risk analysis findings to managers is illustrated in a real case context. This article consists essentially of a report prepared for senior managers of the Nuclear Regulatory Commission to help them make a reactor safety decision. It illustrates the communication of decision analysis findings relating to technical risks, costs, and benefits in support of a major risk management decision: whether or not to require a safety backfit. Its focus is on the needs of decision makers, and it introduces some novel communication devices

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

    International Nuclear Information System (INIS)

    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

  6. Backfitting nuclear power plants to enhance safety, or decommissioning?; Die Sicherheit verbessernde Nachruestungen von Kernkraftwerken oder Stilllegungen?

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Horst

    2010-07-15

    Nuclear power plant operation extensions by up to sixty years are being licensed in Europe and worldwide. In Germany, vociferous clamoring is being repeated to shut down older nuclear power plants. While, as recently as in the 2009 campaign for the federal elections, German plants were considered by many to be the safest in the world, there is now an increasing competition of voices demanding safety enhancement for prolonged operation. This dichotomy of opinions and assessments should be based first and foremost on valid nuclear law in which there is a clearcut system ensuring the safety of nuclear power plants in Germany under the Atomic Energy Act. On the basis of facts and legal principles, the outcome will show whether a nuclear power plant licensed some time ago, and its present major technical safety systems, can be considered safe also from the present vantage point. The safety concept developed by technical and scientific methods constitutes the basis of legal standards which must exclude unsustainable safety conditions. Issues more difficult to solve include the definition of specific identified backfitting measures as against measures further enhancing safety. Legal standardization should be preceded by a broad, transparent and primarily technical public safety discussion, also with international experts and at nuclear power plant locations. (orig.)

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

    International Nuclear Information System (INIS)

    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)

  8. Backfitting the NANO bunkered emergency heat removal system at Beznau

    Energy Technology Data Exchange (ETDEWEB)

    Baschek, Heinz

    1987-09-01

    Construction is due to start in 1988 of bunkered emergency heat removal systems at Beznau nuclear power station. Ten hours core cooling will be maintained by new emergency systems, located in bunkered buildings, in the event of a loss of primary coolant with the primary loop remaining intact.

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

    International Nuclear Information System (INIS)

    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

  10. Solution for backfitting of a controlled atmosphere area in radiopharmaceuticals unit: the mobile unit of decontamination

    International Nuclear Information System (INIS)

    The laboratory of radiopharmaceuticals preparation is a controlled area, defined by good practices of preparation. In order to answer to this regulatory requirement, the environment follow-up is made regularly by the hygiene service of the Grenoble University hospital center (C.H.U.). The results of the last sampling turned out to be wrong. Our objective are to remedy this air contamination and to improve the particulate air quality. A mobile unit of air cleaning allowed to be in accordance with the law at the air cleanliness level for the area of radiopharmaceuticals preparation at the Grenoble C.H.U.. It allows to answer to exigence of good practices of preparation face to the controlled atmosphere areas, in a rapid way, efficient way and cheap way, especially before the reorganization of radiopharmacy place. (N.C.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  13. 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)

    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

  14. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    International Nuclear Information System (INIS)

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter

  15. Operating experience with Beloyarsk fast reactor BN600 NPP

    International Nuclear Information System (INIS)

    The main results of the seventeen-year operation of the BN600 Nuclear Power Plant are considered. The principal backfittings of the main BN600 Power Plant equipment are presented and summarised. (author)

  16. 77 FR 11384 - Removal of Oman from the Restricted Destinations List

    Science.gov (United States)

    2012-02-27

    ... Oman. The Executive Branch recommended, in light of current foreign policy and nonproliferation-related... backfits as defined in 10 CFR Chapter I. VIII. Congressional Review Act Under the Congressional Review...

  17. 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)

    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

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

    International Nuclear Information System (INIS)

    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)

  19. Regulatory analysis for Generic Issue 23: Reactor coolant pump seal failure. Draft report for comment

    International Nuclear Information System (INIS)

    This report presents the regulatory/backfit analysis for Generic Issue 23 (GI-23), 'Reactor Coolant Pump Seal Failure'. A backfit analysis in accordance with 10 CFR 50.109 is presented in Appendix E. The proposed resolution includes quality assurance provisions for reactor coolant pump seals, instrumentation and procedures for monitoring seal performance, and provisions for seal cooling during off-normal plant conditions involving loss of all seal cooling such as station blackout. Research, technical data, and other analyses supporting the resolution of this issue are summarized in the technical findings report (NUREG/CR-4948) and cost/benefit report (NUREG/CR-5167). (author)

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

    International Nuclear Information System (INIS)

    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.)

  1. 78 FR 41835 - Inflation Adjustments to the Price-Anderson Act Financial Protection Regulations

    Science.gov (United States)

    2013-07-12

    ... made the initial changes to the Price-Anderson Act amounts on October 27, 2005 (70 FR 61885), and the first periodic inflation adjustments on September 29, 2008 (73 FR 56451). This final rule makes the... Writing,'' published June 10, 1998 (63 FR 31883). X. Backfit Analysis and Issue Finality The NRC...

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

    International Nuclear Information System (INIS)

    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

  3. PWR power plant reactor maintenance: site experience and technology transfer

    International Nuclear Information System (INIS)

    PWR reactor maintenance activities has considerably expanded during the last few years at Framatome. The services offered by Framatome can be divided into three main categories: - Implementation of backfits aimed at performance and safety improvement and equipment reliability - Technical assistance for plant operators, especially during refuelling - Maintenance and repair services during both scheduled and unscheduled outages

  4. LWR design decision methodology: Phase II. Final report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-01-01

    Techniques were identified to augment existing design process at the component and system level in order to optimize cost and safety between alternative system designs. The method was demonstrated using the Surry Low Pressure Injection System (LPIS). Three possible backfit options were analyzed for the Surry LPIS, assessing the safety level of each option and estimating the acquisition and installation costs for each. (DLC)

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

    International Nuclear Information System (INIS)

    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.)

  6. 76 FR 75771 - Emergency Planning Guidance for Nuclear Power Plants

    Science.gov (United States)

    2011-12-05

    ... Utilization Facilities'' that were published as a final rule in the Federal Register (FR) on November 23, 2011 (76 FR 72560). Additional guidance on one topic not directly related to the EP final rule (i.e... rule discussed that rule's compliance with applicable backfitting provisions (76 FR 72560; November...

  7. NS OTTO HAHN - test report for the period from 5-8 1977 to 13-4 1978

    International Nuclear Information System (INIS)

    Four voyages with freight and one research voyage were made in the period under report. The ship was in the shipyard twice, each time for about 4 weeks. Both times, backfitting work was done and in-service tests were carried out apart from the normal repair and maintenance work. (orig./HP)

  8. Safety upgrade at the Leningrad NPP

    International Nuclear Information System (INIS)

    The LNPP was developed according to the standards of early 70's but, at the same time, during the whole period of operation, the Plant equipment, technological, automatic and control and protection systems were upgraded with regard to changing safety and reliability requirements. Main steps taken during the backfitting stage to improve the reliability and safety of LNPP equipment and systems are discussed

  9. 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.

  10. Development of procedural requirements for life extension of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hun; Son, Moon Kyu [Korea Association for Nuclear Technology, Taejon (Korea, Republic of); Ham, Cheol Hun [The Catholic University of Korea, Seoul (Korea, Republic of); Chang, Keun Sun [Sunmoon Univ., Asan (Korea, Republic of); Paek, Won Phil [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Cheong, Ji Hwan [Baekseok College Cultural Studies, Cheonan (Korea, Republic of)

    2001-03-15

    Current status of regulatory aspects of life extension and upgrading of NPPs is reviewed for major foreign countries. Most countries require similar technical requirements; however, procedural aspects differ country by country. Regulatory systems suitable for NPP life extension is investigated. The procedure and requirements for reassessment of design life should be established first; then it can be incorporated into the PSR system. The concept of 'Current Licensing Basis (CLB)' can be adopted in Korea, but further elaboration for terms and definitions is needed for common understanding between interested groups. The procedure for maintenance and backfitting should also be improved. The Systems, Structures, and Components (SSCs) that require development of regulatory requirements for life extension are identified based on extensive analysis of foreign experiences. By analyzing the rules and regulations related to life extension. Basic directions are suggested to harmonize or establish regulatory systems for life extension, two-step licensing, PSR, and backfitting.

  11. Safety assessment of nuclear power plants equipped with VVER reactors

    International Nuclear Information System (INIS)

    The safety studies of the nuclear generating units of Greifswald and Rheinsberg have produced important basic findings for the assessment of nuclear power plants equipped with VVER reactors. Deficits in engineered safeguards design have been found in the three lines, i.e. VVER-440/V-230, VVER-440/V-213, and VVER-1000. The oldest line is thought to be beyond backfitting, while design deficits in the two other lines can largely be corrected by backfitting measures. In Eastern Europe, safety studies are conducted by GRS and IPSN in close cooperation with national authorities. This is demonstrated by the safety assessment of units 1 to 4 on the Kosloduj site. Studies performed on a national level are supported by the German Federal Ministry for the Environment and, internationally, by the European Community. (orig.)

  12. Development and enhancement potentials of Eastern design-type reactors

    International Nuclear Information System (INIS)

    There are approximately 125 nuclear power plants with Eastern design-type reactors in operation, under construction, or shut down in the countries of central and eastern Europe and in the CIS. Their backfitting is financed by worldwide support at a cost of currently 1.5 billion Deutschmarks. Enhancement activities performed in Russia concentrate on the three major designs, PWR reactors (WWER), breeder reactors (BN), and channel-cooled reactors (RBMK), in order to achieve improved designs for future construction of new plant to replace existing ones. The planning activities for new construction got as far as establishing outline plans, and there are only six more concrete plans providing for new construction of six nuclear power plants based on existing designs, with backfitting requirements to be met for engineered safety. (orig.)

  13. Passive systems for light water reactors

    International Nuclear Information System (INIS)

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

  14. Marginal integration $M-$estimators for additive models

    OpenAIRE

    Boente, Graciela; Martinez, Alejandra

    2015-01-01

    Additive regression models have a long history in multivariate nonparametric regression. They provide a model in which each regression function depends only on a single explanatory variable allowing to obtain estimators at the optimal univariate rate. Beyond backfitting, marginal integration is a common procedure to estimate each component. In this paper, we propose a robust estimator of the additive components which combines local polynomials on the component to be estimated and marginal int...

  15. District heat generating plants - present planning and significant results

    International Nuclear Information System (INIS)

    Combined generation of electricity and heat is a must for economical generation of district heat for the base load range with a minimum environmental impact. More sophisticated desings are taking over both for heat extraction from condensing turbine power plants as well as for backpressure turbine power plants. More and more existing power plants are being backfitted for heat extraction. The advantages and disadvantages of the different concepts are illustrated. The possibilities for nuclear district heat generation are also touched on. (orig.)

  16. Safety enhancement at Beznau NPP

    International Nuclear Information System (INIS)

    The two units of the Beznau Nuclear Power Plant, Switzerland, are presented, and their safety related progress is evaluated. The largest safety enhancement has been the addition of a completely self-contained emergency system. Safety enhancements through backfitting measures in older nuclear power plants, however, have distinctive disadvantages compared to more modern plants. At Beznau NPP, safety has always priority over economics. (N.T.)

  17. Emergency cooling system for the core of a reactor pressure vessel

    International Nuclear Information System (INIS)

    In order to improve the spray distribution in an emergency cooling system for a BWR, the spray nozzles are situated vertically in bores of the pressure containment lid domed towards the inside. The distribution system is therefore situated above the lid and is supported on it. The penetrations for the incoming pipes are situated in the lid. This emergency cooling system is easy to mount and can be backfitted in existing plant. (orig./HP)

  18. Bohunice V-1 and V-2 approach for achieving high availability, reliability and safety

    International Nuclear Information System (INIS)

    Long term operating experience of Bohunice units maintenance activities are overviewed in the paper. Based on common experience of WWER NPP operators, separate maintenance department was established at Bohunice NPP in very early stage of plant operation. Maintenance management, maintenance planning, outage management, diagnostics and monitoring, inspection technologies and backfitting activities are described particularly to demonstrate the capability of Bohunice maintenance department for most complex repairs and maintenance works of nuclear power plant components and equipment, including reactor and turbine itself. (author)

  19. Recent chemical engineering requirements as the result of TMI on-site experience

    International Nuclear Information System (INIS)

    From the experiences gained from the on-site experience at TMI, it is apparent that the role of chemical engineers should increase in order for the nuclear option to proceed in a safe and efficient fashion. It is also obvious that as the results of the reports investigating the causes and effects of the accident come to light and attempts to backfit system designs to prevent a recurrence are studied, more technical demands will be placed on the profession

  20. Ergonomics: an aid to system design

    International Nuclear Information System (INIS)

    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)

  1. FRAMATOME nuclear services

    International Nuclear Information System (INIS)

    FRAMATOME is a French company whose main activities since 1958 have been the design and manufacture of standardized PWR Nuclear Steam Supply Systems. FRAMATOME builds the Reactor Coolant System components and installs and starts-up the extended Nuclear Steam Supply Systems. In addition to the supply of spare parts of tooling, the services offered by Framatome are implementation of backfits aimed at performance and safety improvement and equipment reliability, technical assistance and, maintenance and repair services

  2. Nuclear power plant life management: strategy for long term operation of the Beznau NPP unit 1 and 2

    International Nuclear Information System (INIS)

    The strategy for attaining long-term operation (LTO) of the Beznau nuclear power plants (NPPs) (2 Units) is given. The requirements, technical evaluations for LTO, in addition to considerations for fuel, radwaste disposal, staff and materials management and economic factors, are described. It is shown that, thanks to optimum management strategies, including backfitting and operational improvements, there are no technical reasons to prevent LTO. (author)

  3. Westinghouse computer-based operator support systems

    International Nuclear Information System (INIS)

    Modern plant computers provide access to a large number of plant measured and calculated data. These data can be used by a number of application programs or systems to support the operator's work in the control room. This paper provides an overview of three such systems developed by Westinghouse: an Advanced Alarm Management System (AWARE), a Revolutionary Core Monitoring Program (BEACON), and a Computerized Procedures Package (COMPRO). The AWARE alarm management system was originally designed to be part of the Advanced Control Room design for future nuclear plants. It has been tailored for backfit to operating plants and can be installed during a single plant outage. One of the design goals of the system is to have no alarm in the control room following a reactor trip, as long as the systems are behaving as designed. AWARE can be installed as a full backfit or as a partial backfit using existing annunciator tiles, or provide separate alarm treatment to the plant computer. The BEACON system is a core monitoring, analysis, and prediction tool that provides unparalleled power for understanding and planning core operation. BEACON calculates three-dimensional core power distributions on-line, using an advanced core model. The core model is always in agreement with actual operation because it is continuously updated with measurements obtained from conventional plant instrumentation. Power distribution information is visualized through advanced color graphics to provide users with immediate feedback and comprehensive understanding of core behaviour. The COMPRO system supports the operator in the application of the plant procedures. COMPRO utilizes the written procedures as the basis for textual displays, prompts the operator to the actions to take and provides relevant information about the state to plant parameters. This system guides the user step by step through the plant procedures by monitoring the appropriate plant data and by identifying the recommended course of

  4. Control room systems design for nuclear power plants

    International Nuclear Information System (INIS)

    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

  5. Use of digital photography for power plant retrofits

    Energy Technology Data Exchange (ETDEWEB)

    Kamba, J.J. [Sargent and Lundy, Chicago, IL (United States)

    1995-09-01

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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 H2 concentration inside the confinement. For these tasks systems for confinement atmosphere control are presented, capable of: handling high H2 production rates and cleaning of high contaminated confinement atmosphere. (author)

  8. Working group 5: Safety

    International Nuclear Information System (INIS)

    The technical aspects of safety for the LWR nuclear power plants, and a reprocessing plant are considered. The origin, the type and the extent of the risks for the civil populations are presented for normal working as well as accidental conditions. A general estimate of comparative risks is given for the nuclear industry with respect to other activities. The legal Belgian aspects and their applications, the kind and the quality of the technical testings, the back-fitting of plants are analysed. Considerations are given on the probabilistic analysis, the safety, and the off-shore power plants. (A.F.)

  9. Measures for noise pollution abatement in existing cooling tower systems; Massnahmen zur Geraeuschminderung an bestehenden Kuehlturmanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Niessen, R. [Sulzer-Escher Wyss GmbH, Lindau (Germany)

    1998-09-01

    The operator`s order discussed by the paper was for planning and performance of backfitting measures for noise pollution abatement in an existing cooling tower system equipped with sound attenuation devices. Although the existing plant was operating in compliance with the legal noise emission limits, residents of neighbouring dwellings had been complaining about noise pollution. (orig./CB) [Deutsch] Die Aufgabe, eine nachtraegliche Massnahme zur Laermminderung an einer bestehenden, mit Schalldaempfern ausgeruesteten Anlage zu planen und durchzufuehren, wurde vom Betreiber einer Rueckkuehlanlage gestellt. Der vom Gesetzgeber definierte Grenzwert fuer den Nachtbetrieb wurde mit der bestehenden Anlage zwar erreicht, doch die Anwohner fuehlten eine Belaestigung durch den Anlagenbetrieb. (orig./GL)

  10. Advantages of retrofitting high velocity separators to LWR turbines; experience in VVR NPP Loviisa

    International Nuclear Information System (INIS)

    Erosion-corrosion by wet steam is a concern for VVER operators and also, in numerous LWR power plants of western technology. The backfitting of moisture separators at the HP Turbine outlets is a way to avoid maintenance costs, repairs, replacement of pipes or equipments. Installation of HVS at LOVIISA confirms that this device, whose installation work is reduced to a minimum, is able to remove quite all the water from the steam just a few meters downstream the HP cylinder. A long term operation can be expected for carbon steel equipments, even those previously damaged by erosion-corrosion. (authors). 6 figs., 2 tabs

  11. Advances in Canadian regulatory practice

    International Nuclear Information System (INIS)

    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

  12. Safety of existing installations under dynamic loads: observations on nonlinear response of piping systems - experiments, numerical analyses

    International Nuclear Information System (INIS)

    The nonlinear response of piping systems under base excitation or due to pressure waves caused by simulated breaks and valve closure has been investigated experimentally at the HDR reactor. Structural analysis of ruptured piping and the related design of pipe whips restraints are usually performed on the basis of nonlinear material behavior, with powerful computational techniques being used increasingly. Some aspects of these developments (high-level earthquake tests, high-level pressure wave tests, pipe rupture nonlinear analyses) are summarized with implications for qualification and optimal backfitting of operating nuclear power plants. (Z.S.) 7 refs

  13. Efficient Quantile Estimation for Functional-Coefficient Partially Linear Regression Models

    Institute of Scientific and Technical Information of China (English)

    Zhangong ZHOU; Rong JIANG; Weimin QIAN

    2011-01-01

    The quantile estimation methods are proposed for functional-coefficient partially linear regression (FCPLR) model by combining nonparametric and functional-coefficient regression (FCR) model.The local linear scheme and the integrated method are used to obtain local quantile estimators of all unknown functions in the FCPLR model.These resulting estimators are asymptotically normal,but each of them has big variance.To reduce variances of these quantile estimators,the one-step backfitting technique is used to obtain the efficient quantile estimators of all unknown functions,and their asymptotic normalities are derived.Two simulated examples are carried out to illustrate the proposed estimation methodology.

  14. EFFICIENT ESTIMATION OF FUNCTIONAL-COEFFICIENT REGRESSION MODELS WITH DIFFERENT SMOOTHING VARIABLES

    Institute of Scientific and Technical Information of China (English)

    Zhang Riquan; Li Guoying

    2008-01-01

    In this article, a procedure for estimating the coefficient functions on the functional-coefficient regression models with different smoothing variables in different co-efficient functions is defined. First step, by the local linear technique and the averaged method, the initial estimates of the coefficient functions are given. Second step, based on the initial estimates, the efficient estimates of the coefficient functions are proposed by a one-step back-fitting procedure. The efficient estimators share the same asymptotic normalities as the local linear estimators for the functional-coefficient models with a single smoothing variable in different functions. Two simulated examples show that the procedure is effective.

  15. Recent Nuclear Power Plant Control and Instrumentation activities in the German Democratic Republic

    International Nuclear Information System (INIS)

    In the GDR NPP's with Soviet WWER-type reactors are in operation and in commissioning. Their control and instrumentation systems have been delivered by the Soviet Union. In order to enhance the level on nuclear safety a few additional equipments basing on domestic hardware have been installed, e.g. system for early failure detection. At time backfitting of the control and instrumentation systems of the units 1 to 4 of Greifswald NPP is in preparation. In recent years a computerized training simulator for WWER-440 NPP's has been developed. (author). 2 refs, 1 tab

  16. Consequences for safety and radiation protection of nuclear power plants resulting from the incident in the American TMI-2 nuclear power plant near Harrisburg on March 28, 1979

    International Nuclear Information System (INIS)

    The examinations of German nuclear power plants carried out on the occassion of the TMI-2 incident resulted in backfitting measures for the nuclear power plants that are in operation as well as in the construction or the design stage. Progress has been made with respect to introducing a remote-surveillance system operated by all of the surveillance bodies in the FRG according to the same general specification. The Federal Government has taken initiatives for an international evaluation of the consequences to be drawn from the incident. (orig./HP)

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

    International Nuclear Information System (INIS)

    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

  18. Deterministic and probabilistic methods to assess the safety level of operating nuclear power plants

    International Nuclear Information System (INIS)

    The German safety concept for nuclear power plants gives priority to the deterministic approach, i.e. deterministic analysis and good engineering judgement are primary tools of design evaluation. Probabilistic safety assessment is seen as a supplementary tool to the deterministic approach which provides quantitative information on the occurrence of incidents and thus can be used to check deterministic design assumptions, to evaluate desired plant and system modifications, to optimize backfitting measures and to quantify existing safety margins of operating nuclear power plants, e.g. in the frame of periodic safety reviews.(author)

  19. Model averaging for semiparametric additive partial linear models

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    To improve the prediction accuracy of semiparametric additive partial linear models(APLM) and the coverage probability of confidence intervals of the parameters of interest,we explore a focused information criterion for model selection among ALPM after we estimate the nonparametric functions by the polynomial spline smoothing,and introduce a general model average estimator.The major advantage of the proposed procedures is that iterative backfitting implementation is avoided,which thus results in gains in computational simplicity.The resulting estimators are shown to be asymptotically normal.A simulation study and a real data analysis are presented for illustrations.

  20. Risk - Informed decision making at Loviisa NPP

    International Nuclear Information System (INIS)

    PSA has been used in many ways for risk-informed decision making at Loviisa power station. The most fruitful areas so far include: 1) Identification of dominating risk contributors and possible means for reducing risk by plant modification and improved procedures. 2) Providing risk perspective and economic criteria for assessing backfitting proposals. 3) Assessing the significance of ageing and needs for renewals. 4) Limiting, prioritising and optimising plant modifications. 5) Reducing testing requirements. 6) Justification of temporary as well as permanent configurations and extended outage times. 7) Planning and prioritisation of training programs. (au)

  1. Experience in treatment of components, waste treatment, packaging and shipping related to decommissioning

    International Nuclear Information System (INIS)

    Common practice of the treatment of large amounts of activated core components, contaminated components and waste treatment is described. Single tasks of actual repair and backfitting work comprise up to 800 tons of material to be removed, packed, shipped and decontaminated. Decontamination for unrestricted release has already been successfully performed. Special container systems for shipment and storage of every kind of radioactive waste are presented. Containers resulting in maximum shipping, storage and disposal capacity as well as shipping and storage casks for canned and uncanned fuel are described. The experience gained with nuclear facilities in Europe is correlated to the decommissioning tasks to be performed in the near future. 7 figures

  2. Oyster Creek augmented offgas system startup

    International Nuclear Information System (INIS)

    Oyster Creek Nuclear Station was one of several boiling water reactors requiring backfit with an augmented offgas processing system. Engineering studies conducted during the first half of 1973 resulted in a performance specification for a system with a decontamination factor of 150. The system selected utilized catalytic recombiners with refrigerated charcoal adsorber tanks. Features of the system, up to the startup effort (April of 1977), are described. The various tests established to evaluate system performance under all modes of operation are described along with the results of the completed tests and status of the test program

  3. Use of digital photography for power plant retrofits

    International Nuclear Information System (INIS)

    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

  4. 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)

    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

  5. Experience in using new safety I and C systems at the Beznau nuclear power station; Erfahrungen mit neuer Sicherheitsleittechnik im Kernkraftwerk Beznau

    Energy Technology Data Exchange (ETDEWEB)

    Farruggio, David; Hangartner, Christian; Schaeuble, Thomas [Nordostschweizerische Kraftwerke AG, Doettingen (Switzerland). Kernkraftwerk Beznau

    2009-01-15

    Beznau Nuclear Power Station is made up of 2 nearly identical units built between 1966 and 1972. Unit 1 started commercial operation in December 1969, unit 2, in March 1972. In an effort to always keep the power plant at the latest state of engineered safeguards, backfitting was started early on, also in the field of electrical engineering and I and C. The equipment originally installed for the reactor protection and control system, due to its age, suffered from a lack of support by the vendor and from bottlenecks in spare parts supplies. Consequently, there had to be a change. Planning initiated replacement in 1994. In a first phase, the concepts almost exclusively based on digital control systems were examined. Two of these concepts were worked out in detail in another phase, finally resulting in the decision to implement backfitting of the reactor protection and control system with TELEPERM XS. The reactor protection and control system was replaced in 2000 and 2001. The experience since accumulated has been mainly positive. The hardware is stable in operation, with hardly any failures. The robust architecture prevents the few failures from impacting plant operation. The software has been implemented in such a way that technical process functions are carried out according to design, both in normal operation and during transients. (orig.)

  6. Experience in using new safety I and C systems at the Beznau nuclear power station

    International Nuclear Information System (INIS)

    Beznau Nuclear Power Station is made up of 2 nearly identical units built between 1966 and 1972. Unit 1 started commercial operation in December 1969, unit 2, in March 1972. In an effort to always keep the power plant at the latest state of engineered safeguards, backfitting was started early on, also in the field of electrical engineering and I and C. The equipment originally installed for the reactor protection and control system, due to its age, suffered from a lack of support by the vendor and from bottlenecks in spare parts supplies. Consequently, there had to be a change. Planning initiated replacement in 1994. In a first phase, the concepts almost exclusively based on digital control systems were examined. Two of these concepts were worked out in detail in another phase, finally resulting in the decision to implement backfitting of the reactor protection and control system with TELEPERM XS. The reactor protection and control system was replaced in 2000 and 2001. The experience since accumulated has been mainly positive. The hardware is stable in operation, with hardly any failures. The robust architecture prevents the few failures from impacting plant operation. The software has been implemented in such a way that technical process functions are carried out according to design, both in normal operation and during transients. (orig.)

  7. Instrumentation and control system upgrade plan for operating PWR plants in Japan

    International Nuclear Information System (INIS)

    Digital technology has been applied to all non-safety grade instrumentation and control (I ampersand C) systems in the latest Japanese PWR plants, and has achieved more reliable and operable systems, easier maintenance and cable reductions. In the next stage APWR plants, the digital technology will be also applied to all the I ampersand C systems including safety grade systems. Parallel to the above efforts, many backfitting programs in which the digital technology is applied to operating plants are under way to improve reliability and operability. The backfitting programs for operating plants are proceeded in two phases, synthesizing various utility's needs to improve plant availability and operability, improvement of digital technology, and complexity of the practicable replacement procedures. Phase 1 is a partial application of digital technology, while Phase 2 is a complete application of digital technology. Phase 1 has been implemented in a number of operation plants, while Phase 2 studies are in the design stage, but have not been implemented at this point. This paper presents examples of the partial application of digital technology to operating plants, and the contents of basic design for the complete application of digital technology

  8. Comparative analysis of maintenance cost of PreussenElektra or Sydkraft nuclear power plants

    International Nuclear Information System (INIS)

    The analysis of the PreussenElektra (PE) cost structure for maintenance work shows that the backfitting measures consumed an important share. This is quite obvious for older plants which required substantial backfitting in order to meet the latest KTA safety standards. PE has always been prepared to pay for retrofitting measures whenever progress in safety research and engineering had been revealing deficits in their plant, even if these were below the risk level. However, a technical standard not fully up to KTA safety codes does not in any case mean a serious safety deficit. Instead of insisting on rigid application of regulatory provisions, a problem-oriented approval and licensing policy is found to be much more effective, leaving more room for the owner of the plant to decide and carry out retrofitting measures tailored to his means and under his responsibility. As the plants of PE have been retrofitted in the past few years to comply with the latest safety standards of the regulatory codes, maintenance cost in forthcoming years is expected to decline. As to the Sydkraft maintenance costs, these may take a steep rise, due to the harmonisation policy within Europe, intending to institute the restrictive German regulatory codes as an EU-wide standard and legal basis. Maintenance work therefore has to be very carefully planned for Sydkraft plants. (orig./HP)

  9. Outage planning: basics are the key to success

    International Nuclear Information System (INIS)

    Northeast Utilities Service Company (NUSCO) services approx.1,126,000 customers in Connecticut and western Massachusetts. The company's operating nuclear facilities include Connecticut Yankee, Millstone Unit 1, and Millstone Unit 2. Although NUSCO's earliest commercial operation of a nuclear plant dates back to January 1968, the cost and scheduling section was not created until 1979. The section was created as a corporate service organization composed of four major units: cost engineering, estimating, backfit/betterment planning, and outage planning. To date, 12 of the 28 nuclear refueling outages have been planned and scheduled by the department. There is now a wealth of outage information available from both company personnel and a historical computer database. Current directions within cost and scheduling have been back to basics. The backfit/betterment planning unit has been working with the in-house engineering and design groups to define their work scopes better, to develop a generic work breakdown structure, and to provide better resource planning. The estimating and cost engineering units are working to utilize this to provide true cost and schedule integration. Outage planning will wait for these developments on projects before discussing possible applications in the field

  10. 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)

    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

  11. Balancing safety and economics

    International Nuclear Information System (INIS)

    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.)

  12. Development of transmitter with hybrid-IC for Post-accident monitoring instrumentation

    International Nuclear Information System (INIS)

    After the TMI accident, Post-accident Monitoring (PAM) Instrumentation based on the U.S. Guideline (R.G.1.97) was applied to Japanese PWRs. And we have back-fitted the PAM Instrumentation to old plants step by step. Recently, new type transmitters arrive on the market. They have better accuracy, and stability than old type. However, they cannot be applied as the PAM instrumentation, because new type are insufficient in a qualification for the PAM instrumentation and a modification to endure in-containment accident conditions. Hence, Japanese PWR utilities and Mitsubishi Heavy Industries are developing a new type transmitter for PAM instrumentation to improve accuracy and stability in the period of 1994 through 1996. This paper describes nowadays results in this development of a new PAM transmitter. (author). 8 figs, 3 tabs

  13. Semi-parametric regression: Efficiency gains from modeling the nonparametric part

    CERN Document Server

    Yu, Kyusang; Park, Byeong U; 10.3150/10-BEJ296

    2011-01-01

    It is widely admitted that structured nonparametric modeling that circumvents the curse of dimensionality is important in nonparametric estimation. In this paper we show that the same holds for semi-parametric estimation. We argue that estimation of the parametric component of a semi-parametric model can be improved essentially when more structure is put into the nonparametric part of the model. We illustrate this for the partially linear model, and investigate efficiency gains when the nonparametric part of the model has an additive structure. We present the semi-parametric Fisher information bound for estimating the parametric part of the partially linear additive model and provide semi-parametric efficient estimators for which we use a smooth backfitting technique to deal with the additive nonparametric part. We also present the finite sample performances of the proposed estimators and analyze Boston housing data as an illustration.

  14. Upgrade of process information systems in NPPs, a first step to an overall I and C modernization

    International Nuclear Information System (INIS)

    For power plant I and C-systems an average system innovation cycle of 15 years and a system utilization time of about 25 years is usual. The exception is the computer based information system due to its closer dependency on computer science progress. Economical and operational aspects force utilities to replace computer systems or at least subsystems after 10 to 15 years of operation. Comparing lifespan of further I and C-systems it is obvious to integrate replacement of the plant computer system into an overall I and C-modernization strategy. This is one of the reasons, why backfitting of process information system is more than realizing computerized standard functions like alarm annunciation, logs, archiving and graphics on the basis of modernized hardware and software. 5 figs

  15. Post-licensing imposition of duties, recurrent safety inspection, and nuclear supervision

    International Nuclear Information System (INIS)

    Presently, the only possibility offered by the nuclear law to order and safeguard the backfitting of existing installations to the actual state of the art in science and technology in due time is to use the legal instrument of 'post-licensing imposition of duties', or by way of specific orders of the supervisory authority. The draft amendment now recurs to a former proposal and provides for obligatory, recurrent safety inspections, with plant operators being obligled to produce documents not later than ten years after plant licensing showing an assessment of the plant's engineered safety evaluated by the criteria of the state of the art. Unfortunately this obligation is likely to fail to ensure in practive that the state really comes up to the task of ensuring safe operation of nuclear installations. (orig./HP)

  16. Monitoring radioactivity discharges in containment depressurization

    International Nuclear Information System (INIS)

    In line with the current state of the art in measuring radioactive emissions associated with accidents in nuclear power plants, application of the principle of sample dilution as specified in the KTA accident monitoring criteria is ensured. The main advantages of this concept are the negligible radiation exposure of the plant personnel operating the system and handling the measurement filters; the possibility to use both operating and accident instruments; the smooth transition from normal operation to the accident mode; the ease of backfitting the system; and the lower cost involved in building the system. When detecting transients involving a reduction in activity concentration by several decades, an automatic filter changer should be used, or, when counting aerosol activities, a properly programmed step filter unit should be employed. The definition of maximum filter loads and loading differences in accordance with KTA 1502, para. 4.2.2.1 (4) should be adapted to counting under accident conditions. (orig.)

  17. Nuclear terrorism - an unavoidable companion of nuclear fission?

    International Nuclear Information System (INIS)

    Comparing the security measures provided for with regard to nuclear weapons or to the nuclear inventory of the civilian fuel cycle, it is shown that there are significantly lower standards applied to the storage, processing, and transport of the radioactive fuel material. The difference becomes most obvious when looking at the planning horizons of those responsible for the security measures. The NATO institutions establish their system of security measures on the basis of a dynamical 'threat analysis' reaching far into the future. In the civilian sector, risk analyses and the deduced security measures are well lagging behind the development of realistic risk scenarios. This makes life easier for the operators of nuclear fuel cycle facilities, who otherwise would be obliged to continuously backfit their installations. The cost advantage on the operator's part, however, is obtained at the expense of security. (orig./HSCH)

  18. Corrosion products in the secondary circuit of Beznau NPP

    International Nuclear Information System (INIS)

    The Beznau NPP consists of two 2-loop 380 MWe pressurized water reactors. Unit 1 went into operation in 1969, unit 2 in 1971. Regarding its age, Beznau has to be designated as an old plant. But in fact one has to talk of two young, 'state of the art' units, taking into account the numerous back-fittings. Important measures were the replacements of the steam generators (SGR) realized at unit 1 in 1993 and at unit 2 in 1999. But there were more changes in the secondary systems. Copper was banished from the system completely and replaced by stainless and chromium steel. The condensers were fitted with titanium tubes. New SG with tubes from Inconel 690 TT were installed. Of course the water chemistry was also influenced by these changes. (N.T.)

  19. Beznau upgrades include NANO bunkered system

    International Nuclear Information System (INIS)

    The two 350 MWe Beznau pressurised water reactors (PWRs), are the oldest in Switzerland. Together with the Muehleberg, Goesgen and Leibstadt stations they provide 40% of the total electrical power in the country. In 1979, when the fourth nuclear plant, Leibstadt, was under construction, the Swiss authorities started to evaluate the differences in technology with respect to previous generation plants, and asked Nordostschweizerische Kraftwerke AG (NOK) to upgrade its Beznau units, in line with the latest 'state-of-the-art', as a requirement of operating licence extension. The completion of bunkered safety systems under the NANO project at Beznau 2 on 16 June 1992 and Beznau 1 on 14 July 1993 was a significant accomplishment in which a major safety backfitting programme was implemented without affecting normal plant availability. (Author)

  20. Individual and collective doses in Beznau nuclear power plant

    International Nuclear Information System (INIS)

    The collective dose in Beznau has been in the last 5 years very low: about 210 person-rem per reactor and year or 0.7 rem/GW.y produced energy. This favorable result has been obtained using self-responsible and well trained personnel. As a consequence only limited amount of contractors are working in the plant, the ratio during shut down being about 50% of the total manpower. As a slight drawback, the mean personnel dose of 1200 mrem/y is relatively high, the Ω-value being 1,75. As long as no new information exists on the effect of low level radiation, these mean doses are acceptable. A low collective dose is necessary if the nuclear industry develops. This compromise is similar to that taken to minimize population doses, or back-fitting existing installation, at the cost of doses to professionals

  1. Procedures for maintenance and repairs

    International Nuclear Information System (INIS)

    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)

  2. Introduction and development of good practices at the Barseback nuclear power plant

    International Nuclear Information System (INIS)

    In the pursuit of high availability figures and the increase of reactor safety many different concepts contribute to the final outcome. The Barsebaeck NPP (two identical BWR units 615/615 MW) commenced commercial operation in 1975/1977. The capacity factor has been raised in the course of the years. In 1985 the maximum power level was raised by 6% and a containment venting filter was installed. A gross power production of 100 TWh was reached in October, 1988. This paper describes eight Good Practices which means technology and know-how that have been introduced and developed at our plant with the goal to achieve still better performance and safety level. These good practices and experiences are associated with modifications and backfitting

  3. 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)

    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

  4. Integration of new instrumentation and control systems into existing plant structures

    International Nuclear Information System (INIS)

    Into an existing plant implies new requirements to be met in the process of change. The scope of backfitting, the equipment system, its characteristics, equipment qualification, approval and licensing procedures specific to a country and a set of equipment all affect the expense involved and the duration of the project. Reference is made to reasons for exchange and to a number of points of importance, in the power plant operator's view, in project planning and execution. The following points are treated in particular: - project schedule, - analysis of situation, - project phases: -- scope, technology, interfaces, -- organization of documentation, -- locations of equipment components, -- grounding /shielding, -- heat removal, -- model of zones of protection, -- interface with surveillance computer system, -- access protection (hardware, e.g. limits), -- user interface (software), -- bay test, -- qualification on the basis of international standards, -- configuration and identification documentation, -- simulation, - training. (orig.)

  5. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    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)

  6. Supporting the operators of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    The WWER-440/W-230 reactor lines were designed and developed in the late sixties. Compared to Western plants, their level of technical safety is clearly lower. Backfitting measures designed to upgrade the nuclear safety and reliability of these reactors have been under discussion within the European Community (EC) and the World Association of Nuclear Operators (WANO) for some time. The EC intends to make available some Ecu 100 million (=approx. DM 180 million) over the next three years under PHARE (Pologne, Hongrie Assistance et Reconstruction Economique) assistance program for Eastern Europe within the Environment/Power Economy Area, which amount is to be spent on studies of improving the technical nuclear safety of these reactors. To implement the program, the EC is looking for support to the West European operators of pressurized water reactors and is seeking the cooperation of WANO. (orig.)

  7. Nuclear criticality safety department training implementation

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-06

    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.

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

    International Nuclear Information System (INIS)

    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)

  9. Use of CEDB for PSA

    International Nuclear Information System (INIS)

    The Component Event Data Bank (CEDB) is a centralized bank collecting, at the European level, data describing the operational behaviour of components of Nuclear Power Plants (NPP's) operating in various European countries. It is one of the three event data banks of the European Reliability Data System (ERDS). The CEDB stores information on the operational history (operational times and/or number of demands of intervention in a year, failure-events reports) of components of NPP's well identified by their engineering and operation characteristics. The CEDB (as well as the whole of the ERDS) was conceived as a support to the analyst in his safety assessments for the design of a new NPP or the backfitting of an old one. (orig./HSCH)

  10. MET-RODOS: A comprehensive atmospheric dispersion module

    DEFF Research Database (Denmark)

    Mikkelsen, T.; Thykier-Nielsen, S.; Astrup, P.; Santabárbara, J.M.; Sørensen, J.H.; Rasmussen, A.; Robertson, L.; Ullerstig, A.; Deme, S.; Martens, R.; Bartzis, J.G.; Päsler-Sauer, J.

    the system via on-line connections to on-site local meteorological observations (met-towers and sodars) and via network (either public domain Internet or user-owned point-to-point ISDN) connections to remote national or international meteorological forecasting services. In its final form, scheduled...... for operational use in 1999, the: MET-RODOS meteorological module is intended to service the PODOS system with actual and forecast (+36 h) nuclei-specific air concentrations, deposition values, and gamma radiation estimates on the local, national, and European scale. Provisions are furthermore being...... made for accommodating on line available radiological monitoring data in the meteorological model chains in order for the module to assist with source term determination based on real-time data assimilation and back-fitting procedures....

  11. The KTA nuclear safety rules. Experience with their application shown with the example of an operating nuclear power station, the GKN-1 and GHN-2

    International Nuclear Information System (INIS)

    In case of major modifications being planned in the plant design or operation, - for the purpose of risk prevention and nuclear safety -, the operator has to comply with this existing and applicable regulatory guides and codes in order to provide for the legally embodied preventive means and measures offered by the state of the art in science and technology. The KTA nuclear safety rules may serve in this procedure as an anticipated expert opinion. Care must be taken, however, to assess in detail and for every particular case - on the basis of the given plant overall design - the need for further detailed design specification exceeding the protective goals of the nuclear safety rules. A simply formal approach to an application of safety rules within the process of plant backfitting cannot be accepted by the operators of nuclear installations. (orig.)

  12. Analysis of the role of regulation in the escalation of capital additions costs for nuclear power plants

    International Nuclear Information System (INIS)

    This study examines the role of regulation in the escalation of capital additions costs for nuclear power plants over the past ten years. Unlike previous studies which used a statistical approach to examine the influence of causal factors on the variation in costs, this report is based on actual case studies at four nuclear power plants operated by two utilities. These plants, which are not identified by name, span the entire range of reactor manufacturers. In addition to the evaluation of the role of regulation on capital additions costs, we also examined the contribution of requirements resulting from the accident at Three Mile Island, and where possible, the reasons for utility-initiated backfits. 3 figs., 10 tabs

  13. 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.

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

    International Nuclear Information System (INIS)

    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

  15. Regulatory analysis for the resolution of Generic Issue 130: Essential service water system failures at multi-unit sites

    International Nuclear Information System (INIS)

    The essential service water system (ESWS) is required to provide cooling in nuclear power plants during normal operation and accident conditions. The ESWS typically supports component cooling water heat exchangers, containment spray heat exchangers, high-pressure injection pump oil coolers, emergency diesel generators, and auxiliary building ventilation coolers. Failure of the ESWS function could lead to severe consequences. This report presents the regulatory analysis for GI-130, ''Essential Service Water System Failures at Multi-Unit Sites.'' The risk reduction estimates, cost/benefit analyses, and other insights gained during this effort have shown that implementation of the recommendations will significantly reduce risk and that these improvements are warranted in accordance with the backfit rule, 10 CFR 50.109(a)(3). 19 refs., 16 tabs

  16. Safeguards and retrievability from waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Danker, W.

    1996-05-01

    This report describes issues discussed at a session from the PLutonium Stabilization and Immobilization Workshop related to safeguards and retrievability from waste forms. Throughout the discussion, the group probed the goals of disposition efforts, particularly an understanding of the {open_quotes}spent fuel standard{close_quotes}, since the disposition material form derives from these goals. The group felt strongly that not only the disposition goals but safeguards to meet these goals could affect the material form. Accordingly, the Department was encouraged to explore and apply safeguards as early in the implementation process as possible. It was emphasized that this was particularly true for any planned use of existing facilities. It is much easier to build safeguards approaches into the development of new facilities, than to backfit existing facilities. Accordingly, special safeguards challenges are likely to be encountered, given the cost and schedule advantages offered by use of existing facilities.

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

    International Nuclear Information System (INIS)

    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

  18. Nuclear Criticality Safety Department Qualification Program

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-06

    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.

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

    International Nuclear Information System (INIS)

    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

  20. Probabilistic safety analysis of the Kozloduy NPP units 1-4 (WWER-440/230) using independent emergency feedwater system

    International Nuclear Information System (INIS)

    The safety of the Kozloduy NPP is being promoted by backfitting and improved operational practice. Special measures mitigating potential severe accidents consequences are needed because of some deficiencies in the original design of the four WWER-440 units. In conditions of a total LOCA (Loss Of Coolant Accident) it is impossible to ensure decay heat removal using the existing safety system. In such cases an extra emergency feedwater system independent of the plant's other systems has been introduced which offers a new alternative means of removing the residual heat from the reactor. A probabilistic safety analysis is carried out using the method of event trees. A comparison between the existing safety system and the newly proposed is made. The simulation results of the unit behaviour prove that the damage frequency of the active zone is lower with the new system. 3 refs., 3 tabs., 2 figs

  1. 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.

  2. Nuclear criticality safety department training implementation

    International Nuclear Information System (INIS)

    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

  3. Experience with service and maintenance of a TRIGA Mark II reactor after 24 years of operation

    International Nuclear Information System (INIS)

    The maintenance work and the inspection program carried out at the TRIGA Mark II reactor Vienna after more than two decades of reactor operation is described. With the help of a special underwater telescope all surfaces inside the reactor tank were inspected visually and two beam tubes were inspected with an endoscope. A new water purification loop was installed in 1985, which was followed by a new primary coolant circuit in 1986. The reactor bridge was dismantled, all control rod drives were serviced and some components replaced. As a result of this program it was observed that a TRIGA reactor can be serviced, improved and backfitted even after 24 years of operation with minor efforts. (author)

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

    International Nuclear Information System (INIS)

    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

  5. The CEGB/SSEB response to Recommendations 17 and 18 in the Environment Committee's Report on Radioactive Waste

    International Nuclear Information System (INIS)

    This paper and its accompanying reports respond to recommendation 17 and 18 in the report of the Environment Committee on ''Radioactive Waste'' published in 1986. These recommendations are: ''Recommendation 17: The CEGB and SSEB should conduct a full and published analysis of: (1) the cost of backfitting dry stores to Magnox stations, comparing these costs with the costs of reprocessing vitrification and storage of vitrified high level waste (VHLW); (ii) the feasibility of drying-off Magnox spent fuel once it has been wet in a cooling pond. Recommendation 18: The Department of the Environment should commission and publish a study of the characteristics of Magnox spent fuel in final disposal and of any possible methods of dealing with the particular problems of the long-term chemical instability of uranium metal.'' A feasibility study into the dry storage of Magnox fuel (Rec. 17) and a technical appraisal of the scope for direct disposal (Rec. 18) are included. (author)

  6. NRC, DOE, and industry begin fight for nuclear-licensing reforms

    International Nuclear Information System (INIS)

    New proposals for nuclear licensing must first convince Congress that the reforms can revive the nuclear industry and that over-regulation has been its problem. Congress will give careful scrutiny to the Nuclear Regulatory Commission (NRC) and DOE efforts to streamline licensing to make sure that public health and safety and public access to regulatory decision making are protected. Congress will also challenge whether the industry needs regulatory reform. The NRC and DOE bills pursue the same goals of combining and standardizing construction permits and operating-license procedures to resolve issues in the early stages of a project. The industry sees more incentives in the DOE version, however, because DOE codifies the changes rather than making them discretionary, eliminates a pre-operational hearing and review, and eases backfit requirements. A side-by-side summary of the two proposals compares their provisions for construction permits and operating licenses hearing process, early site approval, and pre-approved designs

  7. RRG: Should certification outrank burden-cutting?

    International Nuclear Information System (INIS)

    This article reports on the recommendations from the Regulatory Review Group (RRG) of the Nuclear Regulatory Commission released on August 20. The general thrust of the report is that the RRG considers the technical substance of the agency's regulations to be either acceptable or under an appropriate review process, but believes that licensees are improperly burdened by arbitrary or vague 'commitments,' decreed by NRC personnel and not clearly based on regulations. The RRG report stated a belief 'that fundamental change can be achieved and maintained only through rulemaking which unambiguously sets the standards to be pursued by both the staff and the industry. If the performance objectives of the regulations are deficient they should be changed and subject to the Backfit Rule. This recommendation is so dominant in effecting change that the Review Group has included as Appendix A a commission paper and proposed rule.'

  8. EPRI/WOG analysis of decay heat removal risk at Point Beach

    International Nuclear Information System (INIS)

    This paper provides a best-estimate probabilistic analysis of Decay Heat Removal (DHR) risk at the Point Beach nuclear power plant. It includes an evaluation of potential plant modifications proposed by Sandia National Laboratories as pare of the NRC's Unresolved Safety Issue program on DHR requirements (USI A-45). The EPRI/WOG analysis yielded a factor of thirty lower core-damage frequency for the sequences included in the scope of the Sandia study. This analysis also yielded a factor of seven reduction in off-site consequences (over and above the core-damage frequency reduction), and estimates of costs that are 50-400% higher for the various proposed backfit modifications. This evaluation, like the Sandia study, concludes that a dedicated SDHR system is not justifiable for Point Beach on a cost-benefit basis

  9. Screening of generic safety issues for license renewal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Faramarzi, A.; Hughes, A.A.; Seth, S.S. (Mitre Corp., McLean, VA (United States))

    1991-12-01

    The US Nuclear Regulatory Commission (NRC) is developing regulations for renewing the operating licenses of nuclear power plants to ensure that they operate safely beyond the present license terms of 40 years. One consideration relates to past resolutions of generic safety issues (GSIs) that did not result in backfit requirements on the licensees. The consideration of an additional operating term of 20 years which the proposed license renewal rule allows, could have retrospective implication for the basis of those GSI resolutions. As part of its technical support to the NRC for the development of license renewal regulations. MITRE has performed an independent review of the GSIs to identify those that could be potentially affected by license renewal considerations. This report describes the screening process and the results of that work.

  10. Nuclear Criticality Safety Department Qualification Program

    International Nuclear Information System (INIS)

    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

  11. Innovations on life management of VVER reactor pressure vessels

    International Nuclear Information System (INIS)

    The embrittlement rate of the pressure vessel weld material is a dominating factor in life management of VVER reactor pressure vessels. In order to maintain adequate safety level, several backfitting measures have been performed in Loviisa. The neutron flux and embrittlement rate was reduced after receiving the first indications of anticipated problems. An increase of emergency core cooling water temperature and other process related changes followed to eliminate and reduce potential transients. Finally, the core weld of Loviisa 1 was successfully annealed in 1996. A current concern is to verify the post-annealing embrittlement rate in order to enable safe and economic life management of the RPV. Post-annealing re-embrittlement is governed by somewhat different mechanisms than the embrittlement of the first irradiation cycle. A new tentative approach for predicting the re-embrittlement rate has been proposed. (orig.)

  12. Optimization of support concepts for piping systems using evolutionary algorithms

    International Nuclear Information System (INIS)

    For the design or backfitting of piping systems a support concept has to be found so that for all loading conditions the stresses and loads on the components, fittings and connections are as low as possible. The problems result from the fact that for different load cases different support concepts are optimal. The load from thermal strains need soft supports in order to avoid constraints, while for seismic loads the optimum support concept needs rigid fittings. Thus, the resulting concept is always a compromise for the load transfer of expected load cases. Based on the example of a bypass cooling loop in a nuclear power plant the authors show that genetic algorithms are adequate tools to solve the optimization problem.

  13. 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)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    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

  15. The safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Nuclear power plant operators and nuclear organizations from the West and from the East cooperate at many levels. The G7 and G24 nations have taken it upon themselves to improve the safety of Eastern nuclear power plants. The European Union has launched support programs, i.e. Technical Assistance to the Commonwealth of Independent States (Tacis) and Pologne-Hangrie: Aide a la Reconstruction Economique (Phare), and founded the European Bank for Reconstruction and Development. The countries of Central and Eastern Europe operate nuclear power plants equipped with VVER-type pressurized water reactors and those equipped with RBMK-type reactors. The safety of these two types of plants is judged very differently. Among the VVER plants, a distinction is made between the older and the more recent 440 MWe lines and the 1000 MWe line. Especially the RBMK plants (Chernobyl-type plants) differ greatly as a function of location and year of construction. Even though they do not meet Western safety standards and at best can be backfitted up to a certain level, it must yet be assumed that they will remain in operation to the end of their projected service lives for economic reasons. (orig.)

  16. Innovative I and C technology products - ensure long-term security of capital investments

    International Nuclear Information System (INIS)

    I and C technology is one of the key systems in nuclear power plants. Keeping it at the latest state of the art is a worthwhile investment for nuclear power plant operators justified by increased safety and higher availability. The TELEPERM registered XS safety I and C technology developed by Framatome ANP, an AREVA and Siemens company, can be employed in various reactor lines made by different manufacturers. The main requirements to be met in safety I and C technology stem from its respective applications. Especially for backfitting purposes, the new system must be easy to integrate into existing concepts and solutions. Moreover, licensability in safety-related applications is an absolute must for any safety system. TELEPERM registered XS allows the broad range to be achieved which is necesssary to meet the safety requirements specific to each plant. In this way, all objectives of safety I and C in nuclear power plants can be met on one uniform systems platform. The use of TELEPERM registered XS involves only minor licensing risks, reduces operating costs, and ensures long-term security of investments. The advantages of a systems platform, such as high functionality and reliability, flexibility in use in a variety of areas, long-term support, extension of the high quantification level of the system, are documented in numerous applications all over the world. (orig.)

  17. Japan: National approach to ageing management

    International Nuclear Information System (INIS)

    In response to the accident at Fukushima Daiichi NPS, the Nuclear Regulation Authority (NRA) was established in September 2012, integrating nuclear regulation functions regarding nuclear safety, security, safeguards, radiation monitoring and radioisotopes, and the Reactor Regulation Act was revised for the purpose of introducing new regulations based on ‘lessons learned’ availability of the latest technical knowledge, as well as trends of overseas regulations, including requirements developed by international organizations such as the International Atomic Energy Agency (IAEA). The main points of the revision include: − Strengthening countermeasures against severe accidents and terrorism; − Back-fitting: NRA can issue an order to comply with new regulatory requirements even to existing nuclear plants without exception; − Limit of the plant lifetime: Up to 40 years one time extension up to 20 years. As of 1st September, 2013, the number of the units in service is 50, 26 units for BWR and 24 units for PWR. 17 units of them are in LTO beyond 30 years, and 3 of them, one BWR and two PWRs, are in LTO beyond 40 years

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

    International Nuclear Information System (INIS)

    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)

  19. 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)

    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)

  20. VVER and RBMK reactors

    International Nuclear Information System (INIS)

    The safety of VVER and RBMK reactors has been discussed a lot after Chernobyl accident. Some improvements have been performed since that especially in RBMK-reactors and extensive programmes for backfitting have been planned and are partly underway. There are two different sizes of VVER reactors, 440 MW and 1000 MW. The design bases and designs itself vary inside the family of two size classes depending on the age of the plant. The oldest VVER-440 is called model 230 and the newest model 213. The oldest VVER-1000 units (two units) are prototypes that have some unique, nonfavorable features. The next stage of VVER-1000 developement (three units) is model V-302 and the remaining 15 plants in operation are model V-320, but even within this latest model there are some differences. The design bases and designs vary also inside the family of the RBMK reactors exactly the same way as in VVERs. The most important design bases of nuclear power plants designed in the former Soviet Union is presented in this paper. Also some safety advantages and disadvantages of these NPPs are discussed. (au). (5 figs.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  4. Potential of low-temperature nuclear heat applications

    International Nuclear Information System (INIS)

    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

  5. Replacement of the reactor protection and control system of the Beznau nuclear power station

    International Nuclear Information System (INIS)

    The Beznau nuclear power station comprises two almost identical pressurized water reactor units; the plant is located in the lower Aare river area. Its aggregate electric power is 760 MW. The plant is operated by Nordostschweizerische Kraftwerke (NOK). In 2000 and 2001, the reactor protection and control systems of both units of the Beznau nuclear power plant were replaced by a modern digital safety instrumentation and control system. That exchange had become necessary because it had become increasingly more difficult to obtain spare parts and maintain the systems. In addition, the expense involved in testing and inspection was to be reduced and fault detection in safety I and C was to be improved. Consistent splitting of the reactor protection system into four physically separate redundancies allowed the safety standard to be raised further, and fire protection to be improved. Continuous backfitting and renewal of the plant in an effort to keep it at the latest level of technical safety has been continued in an impressive way in replacing the old reactor protection and control system by the latest computer-based technology. ''PRESSURE'', the project of replacing the reactor protection and control system, went through several busy phases extending back as far as 1990. Project management was entrusted to a project team of KKB in-house project engineers with expert knowledge in I and C technology and, in addition, detailed knowledge of the processes in the existing plant. (orig.)

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

    International Nuclear Information System (INIS)

    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

  7. Twenty-five years after the foundation of Isar-1

    International Nuclear Information System (INIS)

    The past nineteen years since the commissioning of KKI-1 were characterized by time consuming and very costly backfitting measures designed to keep the plant abreast of current safety requirements. The success of these measures was confirmed by the licensing authority. Also the radiation exposure of the staff and the environment was reduced step by step, and plant availability was raised further. As a consequence, the plant is in a condition now which is better than that of 1977, when the plant was commissioned. KKI-1 thus is well equipped to meet new challenges in the future. The agreement about the electricity directives reached in Luxemburg in mid-1996 will introduce completely new boundary conditions for the power economy in the European Union (EU). According to the new provisions, each EU member country is to open its electricity market to competition step by step from 1999 at the latest. All electricity producers will be able to supply their customers through direct transmission lines. All end user or distributor customers will be in a position to choose their electricity suppliers freely. Also KKI-1 will have to face this competition. The plant is fit for the new era. (orig.)

  8. Probabilistic assessment of dynamic system performance. Part 3

    Energy Technology Data Exchange (ETDEWEB)

    Belhadj, M.

    1993-12-31

    Accurate prediction of dynamic system failure behavior can be important for the reliability and risk analyses of nuclear power plants, as well as for their backfitting to satisfy given constraints on overall system reliability, or optimization of system performance. Global analysis of dynamic systems through investigating the variations in the structure of the attractors of the system and the domains of attraction of these attractors as a function of the system parameters is also important for nuclear technology in order to understand the fault-tolerance as well as the safety margins of the system under consideration and to insure a safe operation of nuclear reactors. Such a global analysis would be particularly relevant to future reactors with inherent or passive safety features that are expected to rely on natural phenomena rather than active components to achieve and maintain safe shutdown. Conventionally, failure and global analysis of dynamic systems necessitate the utilization of different methodologies which have computational limitations on the system size that can be handled. Using a Chapman-Kolmogorov interpretation of system dynamics, a theoretical basis is developed that unifies these methodologies as special cases and which can be used for a comprehensive safety and reliability analysis of dynamic systems.

  9. Continuity and Innovation. 25 years of simulator training for nuclear power plants in Germany

    International Nuclear Information System (INIS)

    The first training simulator for nuclear power plant personnel in Germany was commissioned twenty-five years ago. This date was rather early, both when measured by the development of the German nuclear power program and when compared with the international situation. This farsighted decision demonstrates the importance nuclear power plant operators attach to the sound training of plant personnel. The consistent, and also costly, further development over the past twenty-five years shows that this attitude has not changed. A modern simulator center was built at a total cost of approx. 250 million Euro which can be characterized briefly as follows: - 13 full simulators cover most specific features of existing nuclear power plants. These simulators are backfitted continuously and represent the current state of simulation technology. - Their experience over many years has allowed the staff of approx. 140 to accumulate a high level of know-how in training and simulator operation. Learning from experience is greatly assisted by the fact that all activities are concentrated at one center. - The way in which the center is organized ensures close cooperation with the nuclear power plants responsible for the training of their personnel. - There is a systematic training concept which is being actively developed further. Some of the main developments in recent years include training for emergencies; intensified training in behavioral aspects, such as communication and leadership; the use of simulators for emergency drills; testing of modifications, etc. (orig.)

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

    International Nuclear Information System (INIS)

    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

  11. Problem identification: a key to choosing the right modification

    International Nuclear Information System (INIS)

    In today's economic environment, industry is under ever-increasing pressure to reduce expenditures and maintain or enhance plant availability. Unfortunately, as retained dollars decrease, the funds available to invest in plant improvements decrease. Therefore, the choice of the right backfit project is key to attaining the greatest possible return on investment. This decision-making process is dependent on: (1) identifying the problem correctly, (2) determining the optimum solutions, and (3) evaluating the economic and performance impacts. The identification of problems or improvements that will result in increases to availability, capacity, or productivity must be accomplished in a manner that ensures correct definition and implementation. The availability of a plant can be improved by identifying and resolving problems quickly and correctly. Essential to this is the determination of the root (or underlying) cause. One technique for accurately identifying the root cause is the operations appraisal method, which is the systematic evaluation of data and information to establish the following: (1) problem system, component, or part, (2) problem conditions, (3) apparent cause, (4) root cause, and (5) return on investment. This technique results in discovery of the root cause by tracking the course of events leading up to the problem

  12. Peer evaluation and some valuable lessons

    International Nuclear Information System (INIS)

    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

  13. Duties of a consulting engineer in nuclear technology

    International Nuclear Information System (INIS)

    The duties of a consulting engineer have undergone fundamental changes in the recent past. Their working environment is now characterized mainly by the sufficient availability of electricity in Western Europe in connection with a stagnating economy, the collapse of the Eastern block followed by the opening and intensification, respectively, of the energy and power plant markets between East and West, the decline in national needs to become self-sufficient in energy supply, the declining prices of fossil fuels, the deregulation of the energy market and, specifically in Germany, the change in government. The impact of this change especially on the engineers' profession is explained by the history of Colenco Power Engineering AG. Present activities are summarized as follows: - Ensuring reliable, safe, and low-cost electricity production in existing plants with Western safety standards. - Completion of the power plants under construction in Central and Eastern Europe, with Western safety standards taken into account, and backfitting existing plants. - Decommissioning old nuclear facilities. - Reclaiming old contaminated sites. - Planning and construction of facilities for waste treatment and for interim storage and repository storage. (orig.)

  14. Qualification of a passive catalytic module for hydrogen mitigation

    International Nuclear Information System (INIS)

    The advantage of passive catalytic modules for hydrogen mitigation during core-melt accidents, as compared with active devices like forced-flow recombiners or ignitors, is given by the high reliability of operation and the elimination of potentially violent combustion events. An important step in the qualification of a passive catalytic module system is the determination of the total required capacity and its distribution at various locations in the containment. Experiments and analytic modeling work were performed to qualify the installation of a system of catalytic modules for a large dry pressurized water reactor (PWR) containment. The operational capacity of a prototype catalytic module was determined experimentally, and a corresponding model correlation was developed and integrated into the GOTHIC containment code. This modified code was validated against experimental data. As an application, predictions of the effects, resulting from backfitting a large, dry PWR containment with 50 catalytic modules, were done using the modified code. The catalytic modules keep the hydrogen concentrations below a level of 10% where violent deflagrations could be expected. Local higher concentrations near the release location are inert due to associated low oxygen and high steam concentrations. A proper distribution of the modules in the containment guarantees full mixing of the atmosphere due to natural convective currents

  15. 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)

    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 H2 concentration in the containment during DBA and severe accident situations. The new HMS contains of overall H2-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

  16. Monitoring of water level inside reactor pressure vessel

    International Nuclear Information System (INIS)

    Up to the TMI accident the water level inside the pressurizer was used to monitor the water inventory inside the primary cooling system of pressurized water reactors. The TMI accident showed that this was not a reliable measurement for the reactor coolant inventory inside the reactor pressure vessel. For this reason there was a demand for a measurement of the water level inside the RVP, independent from the existing one inside the pressurizer and with a diverse measuring method. For WWER reactors a new level measurement system was developed to monitor the water level inside the reactor pressure vessel by means of the KNITU, resp. KITU level probe which meet all the mentioned engineered safeguards and geometric and constructive requirements. First backfitting s of the new level measurement system in the WWER s 440 in Bohunice V1 (Slovakia), unit 1 (1998) and unit 2 (2000), Novovoronezh (Russia), unit 4 (1999) and Kola (Russia), unit 1 and unit 2 (1999) show very good operational results. (Authors)

  17. 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.

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

    International Nuclear Information System (INIS)

    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.)

  19. CANDU safety management in Pakistan. A status report

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  1. Simulation of the primary circuit breaks in the training simulator

    International Nuclear Information System (INIS)

    The pressurized water reactors (PWR) have been designed to operate in the way that normally there is only one-phase liquid in the primary circuit. In the primary break accidents, however, the primary pressure decreases as a consequence of the coolant discharge and a two-phase condition may occur. Large computer codes like RELAP5 and TRAC, which consume lots of computer time, are used for the simulation of reactor systems in small break loss of coolant accidents (SBLOCA). The full-scope training simulator of the Loviisa pressurized water reactor has been improved by backfitting a two-phase model SMABRE to the simulator. The two-phase model is necessarily needed when the primary circuit breaks are simulated. The basic physical models in the SMABRE are similar to those in the large simulation codes, but making some simplifications the real time simulation on a PDP-11/70 computer is possible. Two nodalization models, one with 26 volumes and the other with 61 volumes, have been applied in the training simulator. The main reason for the use of the more detailed model is the need to take into account the variations of primary loop elevations. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

  4. Some practical implications of source term reassessment

    International Nuclear Information System (INIS)

    This report provides a brief summary of the current knowledge of severe accident source terms and suggests how this knowledge might be applied to a number of specific aspects of reactor safety. In preparing the report, consideration has been restricted to source term issues relating to light water reactors (LWRs). Consideration has also generally been restricted to the consequences of hypothetical severe accidents rather than their probability of occurrence, although it is recognized that, in the practical application of source term research, it is necessary to take account of probability as well as consequences. The specific areas identified were as follows: Exploration of the new insights that are available into the management of severe accidents; Investigating the impact of source term research on emergency planning and response; Assessing the possibilities which exist in present reactor designs for preventing or mitigating the consequences of severe accidents and how these might be used effectively; Exploring the need for backfitting and assessing the implications of source term research for future designs; and Improving the quantification of the radiological consequences of hypothetical severe accidents for probabilistic safety assessments (PSAs) and informing the public about the realistic risks associated with nuclear power plants. 7 refs

  5. EPRI's zebra mussel monitoring and control guidelines

    International Nuclear Information System (INIS)

    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

  6. Regulatory analysis technical evaluation handbook. Final report

    International Nuclear Information System (INIS)

    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

  7. Use of quantitative safety goals and probabilistic risk assessment in regulatory decision-making

    International Nuclear Information System (INIS)

    The establishment of quantitative safety goals as an expression of acceptable risk level and the use of probabilistic risk assessment (PRA) as a means of estimating level of risk are gaining increased acceptance as a means of rationally improving the regulatory decision-making process. In the USA, the Nuclear Regulatory Commission and the nuclear industry have made significant advances in attempting to apply these tools in practice. This paper presents a review of US nuclear industry proposals for the establishment and use of quantitative safety goals and PRA. The structure and rationale for a set of quantitative safety goals which address (1) individual risk, (2) population risk, (3) cost/benefit criteria for risk reduction, and (4) core melt frequency are presented. In concert with this, a process is described for applying these quantitative safety goals and utilizing PRA studies in determining whether existing regulations and plant designs are adequate for controlling the introduction of new requirements into the regulations for plant-specific backfitting. Suggestions are provided regarding the use of these techniques by developing countries in establishing their regulatory policies and requirements. (author)

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

    International Nuclear Information System (INIS)

    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 LN2 test, Source LH2-H2O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  9. Existing nuclear power plants and new safety requirements - an international survey. A description of the legal situation and of the regulatory practice in eight countries and in Germany

    International Nuclear Information System (INIS)

    In our days, the question of whether existing nuclear power plants can be expected to comply with new standards is relevant for many reasons. The idea of writing this report was sparked by the fact that the German Federal Ministry of the Environment is planning a thorough revision of the regulations concerning nuclear safety. Since in Germany, according to the latest amendment to the Nuclear Act, a licence for a new plant cannot be granted, this project inevitably raises the basic question of whether the existing plants can be forced to comply with new safety regulation, if necessary by performing substantial backfitting. Aim of the enquiry is to find out how the question outlined above - new requirements for existing nuclear power plants - is dealt with in nine countries, namely Germany, Switzerland, France, Sweden, Finland, the United Kingdom, the USA, Spain and Belgium. In order to give a legible and qualified account, the authors have also investigated and depicted the general legislative and regulatory framework for nuclear of each country. Therefore, the book can also be read as a general introduction into the legal system and regulatory practice of these countries. (orig.)

  10. Research on hydrogen risk mitigation resulting from hypothetical severe accident. Major results from the 4th framework programme and actual research in the 5th framework programme

    International Nuclear Information System (INIS)

    The evolution and release of hydrogen in the course of hypothetical severe accidents in light water reactors is seen to be a potential cause of early containment failure due to possible follow-on reactions in the containment. The cause of concern is the volume of hydrogen produced in the zirconium-water interaction of fuel cladding tubes and structural materials in the reactor core. As a result of deflagration processes and transfer processes, respectively, from deflagration to detonation, this hydrogen can cause local pressure peaks which exceed the containment design basis. For this reason, the phenomena of hydrogen generation, distribution, and reactions have been studied worldwide for a number of decades, and concepts are being developed for controlling this situation. In Germany and in other European countries, the concepts pursued comprise the use of recombiners and ignition devices or a combination of both systems. These systems are being planned for backfitting into existing nuclear power plants and for installation in the European Pressurized Water Reactor (EPR) under development. Since the mid-nineties, several research programs have been devoted to this topic within the framework of joint European reactor safety research. Activities are concentrated on hydrogen generation, hydrogen distribution, the effectiveness of recombiners as well as turbulent combustion and the transition from deflagration to detonation, among others. (orig.)

  11. 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.

  12. Experimental drop testing of waste containers for the Konrad repository - 59269

    International Nuclear Information System (INIS)

    Document available in abstract form only. Full text of publication follows: The Konrad repository for not heat generating radioactive wastes was licensed in 2002 primarily. Due to legal actions the final confirmation of this license took place not until 2007. Subsequently, the Federal Office for Radiation Protection (BfS) began scheduling backfitting of the former iron ore mine into a repository. The licensed repository volume is 303, 000 m3 considering estimations of expected waste volumes to be disposed off. The mine itself would offer a much larger volume. Waste packages can be disposed off as recently as the repository is ready for operation what is expected not before the end of this decade. Nevertheless, there is high interest of qualified and certified waste conditioning and packaging for disposal today, what for from BAM and BfS tested, evaluated and certified containers are needed. In recent years numerous container prototypes made of steel, concrete and ductile cast iron have been tested by BAM, the Federal Institute for Materials Research and Testing in Germany. To cover the Konrad test requirements in a conservative manner container drop tests are performed mostly onto the unyielding IAEA target of BAMs large drop test facility instead of a representative foundation of the repository

  13. Westinghouse Electric. Know-how and top technology from Germany support non-polluting, safe, cost-effective power supply worldwide

    International Nuclear Information System (INIS)

    Westinghouse Electric Company LLC is one the world's leading firms in the commercial nuclear power field with a staff of approx. 15,000, of whom approx. 5,000 work in Europe. As part of the Toshiba Group, Westinghouse supports power utilities in the Americas, Asia, and EMEA (Europe, Middle East, Africa) regions with a broad range of products and services in nuclear power plants, nuclear fuel, nuclear services, and nuclear automation. The German-based company, Westinghouse Electric Germany GmbH, has more than 500 persons at the locations of Mannheim; Hamburg; Baden, Switzerland; and Metz, France. For more than 40 years, it has been successfully operating in field services, plant engineering, waste management, and nuclear automation. The Mannheim head office works the nuclear markets in Germany, Switzerland, the Czech Republic, Slovakia, and Hungary. Under global resource utilization and products schemes, staff from Germany is employed also in projects all over the world. Present construction of a large number of new plants of the AP1000 registered reactor line in China and USA as well as planning and licensing steps for the construction of new nuclear power plants in Europe constitute a major contribution by Westinghouse to the worldwide renaissance of nuclear power. As a partner of utilities, Westinghouse also upgrades existing plants by backfitting and modernizing components and systems, management of aging, safety analyses, non-destructive testing, replacement of safety and operations I and C etc. for plant life extension and safe, economically viable continued operation. (orig.)

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

    International Nuclear Information System (INIS)

    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 SnO2 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

  15. Evaluation of BDB accident management in PSA for recent German 1300 MW PWRs (Konvoi)

    International Nuclear Information System (INIS)

    The Siemens AG/KWU has been performing the probabilistic safety assessment (PSA) for the nuclear power plants (NPPs) for more than 25 years for purposes of design optimization, safety research and special licensing issues. Focus of the PSA application nowadays is towards development of advanced NPPs such as EPR and 1,000 MW BWR, periodic safety review of operating plants, development and implementation of BDB (beyond design basis)-AM (accident management) measure, and so on. Here were discussed on the last two topics. As a results, PSA gave underline of high safety level on basic design in a plant expressed by the already low hazard states frequency and the balanced design, and it was recognized that efficiency of the BDB emergency measures and procedures expressed reduction of frequency required for plant damage states, importance of the emergency procedures for mitigating damage potential of reactor coolant pressure boundary failure under pressed conditions, and representation of backfitted BDB AM measures for an additional level in multi-level safety concept of the plants. (G.K.)

  16. PSA Level 2 activities for RBMK reactors

    International Nuclear Information System (INIS)

    Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

  17. Survey of proposed functional requirements for a disturbance analysis and surveillance system

    Energy Technology Data Exchange (ETDEWEB)

    Sides, W.H. Jr.; Oh, C.B.; Knight, P.F.

    1980-10-01

    A program to enhance the capabilities of operators of nuclear power plants is being pursued by the US Nuclear Regulatory Commission (NRC). The program includes improvements in plant monitoring, diagnostic and corrective action aids, operator-process communication, and operator training. Concerning diagnostic aids, a disturbance analysis and surveillance system (DASS) was considered which would monitor the plant for the approach or occurrence of disturbance conditions and would assist the operator in returning the plant to normal operation or to help mitigate the consequences of a failure condition or misoperation. The NRC had requested Oak Ridge National Laboratory to survey the functional requirements being proposed for a DASS. In fulfilling this task, the proposed requirements were categorized according to whether they could be realized in the short term and backfitted to existing plants or whether they could be realized only in the long term by incorporation into new plant designs. In addition, several recommendations concerning DASS development were made for consideration by the NRC. Finally, the effects of human factors on a DASS were evaluated, and the results are discussed in this report. 12 refs., 6 figs.

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

    International Nuclear Information System (INIS)

    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

  19. The SV4 program system

    International Nuclear Information System (INIS)

    The electronic control system of the SV4a triaxial neutron spectrometer in the DIDO reacor was completely revised. Within the framework of the backfitting work, a new concept was established for the control program of the reactor experimental computer. The PDP11 of the former configuration was replaced by an IBM AT03 computer, and the CAMAC bus system was replaced by the SMP bus system (Siemens). An independent, fixed-programmed control unit (S5, Siemens) actuates the motor driving units upon end switching and signalling contacts and thus protects the instrument from destruction by operating errors or computer malfunctions. Now controllable functions to be mentioned are: Sample temperatures, external field parameters represented by voltages, the two tilting angles at sample holder, and the beam limiting diaphragms ahead of and behind the sample. The diaphragms and crystal-tilting devices for adustment of the monochromator and the filters at the monochromator exit likewise were made accessible by automatic, computer-controlled processes. (orig./DG)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  2. Seismic and tsunami safety margin assessment

    International Nuclear Information System (INIS)

    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)

  3. Regulatory aspects of radiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    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

  4. 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.

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

    International Nuclear Information System (INIS)

    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.)

  6. Workshop on environmental qualification of electric equipment

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  8. 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.

  9. Retrofitting of power plants. Chances and partnerships; Kraftwerksmodernisierung. Moeglichkeiten und Partnerschaften

    Energy Technology Data Exchange (ETDEWEB)

    Bald, A. [Siemens AG, Bereich Energieerzeugung (KWU), Erlangen (Germany); Schwegmann, P. [Siemens AG, Bereich Energieerzeugung (KWU), Erlangen (Germany)

    1997-01-01

    Thousands of power plant managers in the former Soviet Union and the COMECON countries were compelled in the last few years to make a virtue of necessity, adopting the slogan that says ``necessity is the mother of invention`` in their efforts to keep their power plants operating, although there was no way of getting spare sparts or assistance from the general planning boards in Moscow; due to their improvisatory skills they accomplished a great achievement and managed to maintain electricity supply to date. However, the resulting situation today is that the great majority of power plants in the former COMECON member states, i. e in the CIS and in Central and East Europe, badly need repair and backfitting. The article discusses chances and potentials of power plant retrofitting activities, cooperative activities on the part of western countries, and financial support programmes. (orig./RHM) [Deutsch] Tausende von Kraftwerksdirektoren in der ehemaligen Sowjetunion und in den RGW-Staaten haben aus der Not eine Tugend machen muessen: Mit Improvisation und Flexibilitaet hielten sie ihre Kraftwerke am Laufen - Hilfe und Ersatzteile waren von den zentralen Planungsstellen in Moskau kaum zu erwarten; es war eine ungeheure Leistung, dass die Stromversorgung ueberhaupt noch aufrechterhalten wurde. Als Ergebnis dieser Zwangslage sind die meisten Kraftwerke im ehemaligen RGW-Raum, dem heutigen Mittel-Osteuropa und der GUS, dringend ueberholungsbeduerftig. Moeglichkeiten der Kraftwerksmodernisierung, Kooperationen bei der Rekonstruktion und Fragen der Finanzierung werden im folgenden erlaeutert. (orig./RHM)

  10. Nuclear reactor safety and Federal regulation

    International Nuclear Information System (INIS)

    Public confidence in nuclear reactors requires that technical people translate complex safety information into a form that the public can understand well enough to make a judgment. An overall picture is drawn of the major areas of concern: (1) risks and safety measures, (2) government regulation, (3) licensing, (4) plant operation, (5) safety experience, and (6) quality assurance. Although the possibilities of a reactor core melting through the concrete containment barrier are slight, rigorous safety efforts are required. Government regulation and technical developments have developed concurrently so that the high standards set for government facilities can be carried over to commercial efforts. There are two stages in the licensing procedure: a construction permit and an operating license. Reviews of the proposed site, design, emergency cooling systems are all held, followed by a public hearing. Inspection and backfitting of new safety equipment are required in operating plants. The 60 plants now in operation have a good performance record, but good management for quality assurance increases safety and efficiency factors

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

    International Nuclear Information System (INIS)

    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 U3O8 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

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

    International Nuclear Information System (INIS)

    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

  13. A posteriori verification and validation of a tritium dispersion and consequence model

    International Nuclear Information System (INIS)

    An American National Standards Institute (ANSI) posteriori (backfit) process, available to provide software quality assurance (SQA) for software developed outside of required qualification protocol, has been applied to the special-purpose, versatile tritium dispersion and consequence model, UFOTRI, a computer model developed at the German Karlsruhe laboratory. UFOTRI was chosen because of its strengths in initial tritium-related consequence analyses and its potential for application in a Department of Energy accident analysis context. The six-task process met key ANSI requirements and was performed during a several-month level of effort. Included project deliverables were Assessment, Test Plan, Configuration Procedure, Error Notification Procedure, Comprehensive Technical Report, and SQA Qualification Report documentation. Comparison to acute release conditions is still in progress, but results to date indicate satisfactory, bounding predictions can be achieved with UFOTRI relative to measurements. Results of this compact effort appear to identify UFOTRI as a suitable candidate for a software toolkit, i.e., minimum verification and validation (V and V) requirements are satisfied, and a configuration controlled version is deemed appropriate for use in a DOE accident analysis context

  14. 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)

    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

  15. Proposal for the extension of ''probabilistic safety analyses'' on internal fires in German nuclear facilities

    International Nuclear Information System (INIS)

    The goal of this examination was to adapt the application of the probabilistic safety analysis in the Federal Republic of Germany to the international state of science and technology especially for the initiated event ''fire''. For that purpose extended investigations were necessary for special questions concerning the methods used in the probabilistic safety analysis for Nuclear Power Plants (NPP). Furthermore the limits of probabilistic safety analysis of fire caused events were to be checked on the basis of existing analyses, and proposals were to be worked out for future proceeding. The result of the investigations was basically an extensive comparison of different methods with assessment on the basis of existing probabilistic fire risk analyses and pertinent German regulations. The recommended procedure describes a starting point for the method to be discussed which utilizes proved analytical procedures and again was checked for applicability. With this method the probabilistic safety assessment of fire protection in NPP's can be done on a unified basis and the valuation of backfitting measures can be extended successfully into the area of fire protection. (orig./HP)

  16. Cost-benefit considerations in regulatory analysis

    International Nuclear Information System (INIS)

    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

  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)

    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)

  18. Surry nuclear power station: 25th anniversary

    International Nuclear Information System (INIS)

    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.)

  19. Experience with safety I and C modernization at Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Modernization of the reactor protection system represents a technical and safety technology backfit of decisive significance within the safety enhancement measures of Paks Nuclear Power Plant. Installation and commissioning of the new reactor protection system was successfully performed in Units 1, 2, 3 and 4 during the annual outages in 1999, 2000, 2001 and 2002, respectively. A very large contract for extensive refurbishment affecting all four reactor units and the associated full-scope simulator was awarded to Siemens AG, now Framatome ANP GmbH, in a two-step tendering process. This project has several features that are significant for TELEPERM XS projects: The four units were to be equipped with a digital reactor protection system (RPS) of the same design based on the process system requirements specified by Paks Rt.; Every year the RPS of one plant unit was to be replaced during a normal extended outage; The units were to be started up immediately after implementation of the new system without on-line open-loop testing. This paper outlines the reasons for the refurbishment, the refurbishment scope, the structure of the reactor protection system, the verification and validation methods used, the planning and implementation of installation and commissioning, as well as special features of the TELEPERM XS system platform. (author)

  20. East-West cooperation in matters of reactor safety

    International Nuclear Information System (INIS)

    In the Commonwealth of Independent States (CIS), i.e. in the former Soviet Union, and in the countries of East Europe, there are currently 63 nuclear power plant units of Russian design in operation, and 18 units are under construction. Several recent studies are available that enlighten the safety level of these plants. According to this body of information, the reactors of the Chernobyl design type as well as the first generation of the WWER reactors (PWRs) ought to be shut down soon, and the recent WWER plants need backfitting for safety reasons. There is little hope that this can be done in the near future, due to serious financial problems and the importance of nuclear energy for power supply in many of these countries. However, numerous international projects and programmes, partnership agreements between German and Eastern nuclear power plant operators, various projects for assistance in the establishment of suitable institutions and authorities, and scientific-technical cooperation projects considerably contribute to improving the situation. (orig.)

  1. Safety issues and their ranking for 'small series' WWER-1000 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    specific plant status. The review of the safety features of 'small series' WWER-1000 plants shows that the main safety concept of these reactors is similar to that of model 320 with respect to the nuclear island arrangement, the amount of safety systems and the main process parameters of the primary and secondary circuits. However, the 'small series' WWER-1000 plants have major deficiencies such as a lack of separation of redundant safety systems and a single set of the reactor protection system for technological parameters which do not meet the current national standards and international practice. Differences in engineering design solutions, quality of manufacture and reliability of equipment have been revealed as deficiencies. About one third of the design safety issues have been identified by operational experience. The majority of safety issues have been identified as deviations from current standards and practices which have evolved since the WWER-1000 NPPs were designed. Much of the backfitting and upgrading work recognized as being required has been or is being performed. This activity was initiated by the WWER Owners Group and since the early 1990s international assistance has played an important role in the process of safety improvement of these NPPs. The current status of plant specific backfitting varies from site to site, depending on national regulatory requirements and the available financial means. The review of the plant specific status also indicates that certain safety issues have already been solved in some units. This report presents the information currently available to the IAEA on safety issues and safety improvement measures in 'small series' WWER-1000 plants. The IAEA intends to update this information regularly and make it available to the interested parties as part of the technical database developed within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs

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

    International Nuclear Information System (INIS)

    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

  3. 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

  4. The power of British Energy

    International Nuclear Information System (INIS)

    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.)

  5. Methods development to evaluate the risk of upgrading to DCS: The human factor

    Energy Technology Data Exchange (ETDEWEB)

    Ostrom, L.T.; Wilhelmsen, C.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-04-01

    The NRC recognizes that a more complete technical basis for understanding and regulating advanced digital technologies in commercial nuclear power plants is needed. A concern is that the introduction of digital safety systems may have an impact on risk. There is currently no standard methodology for measuring digital system reliability. A tool currently used to evaluate NPP risk in analog systems is the probabilistic risk assessment (PRA). The use of this tool to evaluate the digital system risk was considered to be a potential methodology for determining the risk. To test this hypothesis, it was decided to perform a limited PRA on a single dominant accident sequence. However, a review of existing human reliability analysis (HRA) methods showed that they were inadequate to analyze systems utilizing digital technology. A four step process was used to adapt existing HRA methodologies to digital environments and to develop new techniques. The HRA methods were then used to analyze an NPP that had undergone a backfit to digital technology in order to determine, as a first step, whether the methods were effective. The very small-break loss of coolant accident sequence was analyzed to determine whether the upgrade to the Eagle-21 process protection system had an effect on risk. The analysis of the very small-break LOCA documented in the Sequoyah PRA was used as the basis of the analysis. The analysis of the results of the HRA showed that the mean human error probabilities for the Eagle-21 PPS were slightly less than those for the analog system it replaced. One important observation from the analysis is that the operators have increased confidence steming from the better level of control provided by the digital system. The analysis of the PRA results, which included the human error component and the Eagle-21 PPS, disclosed that the reactor protection system had a higher failure rate than the analog system, although the difference was not statistically significant.

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

    International Nuclear Information System (INIS)

    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

  7. Evaluation of severe accident risks and the potential for risk reduction: Peach Bottom, Unit 2. Main report. Draft for comment, February 1987

    International Nuclear Information System (INIS)

    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 I containment (Peach Bottom, Unit 2). 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 modeling of the common-mode failures for the dc power system, the likelihood of offsite power recovery versus time during a station blackout, the probability of drywell failure resulting from meltthrough of the drywell shell, the magnitude of the fission product releases during core-concrete interactions, and the decontamination effectiveness of the reactor enclosure building. Most of the postulated safety options do not appear to be cost effective, although some based on changes to procedures or inexpensive hardware additions may be marginally cost effective. This draft for comment of the SARRP report for Peach Bottom does not include detailed technical appendices, which are still in preparation. The appendices will be issued under separate cover when completed. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150. (author)

  8. Absorber materials in CANDU PHWRs

    International Nuclear Information System (INIS)

    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 the relatively benign environment of low pressure, low temperature heavy water between neighbouring rows or 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 redesigned back-fit resolved the problem. (author). 3 refs, 8

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

    International Nuclear Information System (INIS)

    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

  10. Post-licensing orders under atomic energy law. With special emphasis on provisions for accident management in nuclear installations; Nachtraegliche Anordnungen im Atomrecht. Unter besonderer Beruecksichtigung der Massnahmen des sogenannten anlageninternen Notfallschutzes

    Energy Technology Data Exchange (ETDEWEB)

    Gemmeke, R.

    1995-12-31

    The author discusses the aspects of implementation of backfitting measures by way of a so-called informal order of an administrative authority based on retrofitting agreements between the authority and the facility operator. The author`s review of the legal aspects leads to the conclusion that this procedure is admissible also in the case of facility operators agreeing to implement retrofitting measures beyond the scope prescribed by existing law. It is explained that the legal provisions for the prevention of damage cover not only measures for prevention of defined risk or damage, but also risks or damage caused by probable risks beyond the danger threshold expressly defined, while such orders of the administrative authority are not admissible in the procedure for licensing, or during the operational phase of a given installation (protection of third parties). The author qualifies post-licensing orders as an interference with the property rights of facility owners protected by the constitution, as in his eyes the provisions of the atomic energy law comply with the constitutional requirements of protection of existing rights. The book further discusses the relationship between post-licensing orders and subsequent, amending licences, and presents considerations on the planned revision of the Atomic Energy Act. (HP) [Deutsch] Der Autor widmet sich der Realisierung von Nachruestungsmassnahmen im Wege des wg. informalen Verwaltungshandelns durch Sanierungsabsprachen Behoerde-Betreiber. Diese seien auch dann zulaessig, wenn die Betreiber sich fuer Modernisierungsmassnahmen bereiterklaeren, die ueber das gesetzlich Gebotene hinausgehen. - Die atomrechtliche Schadensvorsorge umfasse neben der eigentlichen Gefahrenabwehr auch die der Gefahrenschwelle vorgelagerte Risikovorsorge, waehrend beim sog. Restrisiko bei der Genehmigungserteilung und waehrend der Betriebsphase einschraenkende behoerdliche Anordnungen nicht zulaessig seien (Drittschutzproblematik). - Der Autor

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

    International Nuclear Information System (INIS)

    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)

  12. Safer design for a nuclear power plant

    International Nuclear Information System (INIS)

    During the regulatory process for the issuing of the construction permit and the operating licence of the first Austrian nuclear power plant, more than 1200 injunctions have been issued for increasing its safety standard. In principle they belong to three groups: quality assurance and quality control; the improvement of the design; and probabilistic issues. Examples of all these three groups are given. When discussions with the parties in the regulatory process on the issuing of the operating licence were going on, work at the nuclear power plant was suddenly terminated following the negative outcome of a referendum. The main content of the discussions was that the nuclear inspectors keep permanent control over the plant and have a permanent record of occurrences there, that participation of the regulatory body is included in all issues which might influence the safety standard of the plant, and that the regulatory body may issue new injunctions on the operation of the plant if new standards arise from backfitting ensuing from lessons learned, from the treatment of generic issues, from new rules and regulations and from reactor safety research. Special attention is given to the process of mothballing the plant as was necessary after the referendum. The work on the plant was terminated in an orderly way; a final report was issued which stated what still would have to be done at the plant in order to go into operation. The mothballing began by demounting some systems, emptying others and shutting down a third group. Some ventilation systems are in operation. These activities are also recorded in reports; these, together with a final report of the status reached, could be the basis for revitalization work. Finally it is shown how Austria, with its limited means in terms of funds and personnel, is dealing with the problems of keeping the safety standard of the plant as high as at the plants in other countries with more funds and personnel available. (author)

  13. 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.

  14. Designing for nuclear power plant maintainability and operability

    International Nuclear Information System (INIS)

    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)

  15. Nuclear energy: state of the art, necessity and acceptance, possible developments

    International Nuclear Information System (INIS)

    Nuclear energy is a relatively young, in many countries well-established technology. The operational records of commercial plants vary between satisfactory and excellent. Numerous incidents and a few accidents have been reported, which, however, have not demonstrably led to lethal cases prior to the Chernobyl accident. The environmental impact is small: Radiation for non-professionally exposed persons well below the natural background, no greenhouse-gases. The quantities of (highly) radioactive waste accumulated to date are small, but have to be safely stored for incredibly long times; the origin of the problems is rather of intellectual than of insoluble technical nature. The safety standard of western plants is high, but is based on fast acting safety measures and systems, which make nuclear plants complex and less 'forgiving' against failures. On the other hand, eastern European nuclear plants show considerable safety deficits, which need to be overcome. The lack of acceptance for nuclear energy is partially caused by the risk profile of the plants operated today: This profile results into an extremely low frequency of occurence of catastrophic events, without, however, excluding them. The current risk profile can be influenced by technical means. Corresponding technical developments are graduated in time and cover the whole domain between backfitting of existing plants over evolutionary designs up to radical changes for far-future concepts; for the latter the goal of a more far-reaching elimination of severe accidents and/or catastrophic radioactive releases is aimed at, by means of an increased use of passive systems and inherent safety features. (author) 10 figs., 4 tabs., 30 refs

  16. OECD/NEA International Common Cause Failure Data Exchange (ICDE) project - insights and lessons learnt

    International Nuclear Information System (INIS)

    Events initiated by common-cause-failure (CCF) can significantly affect the availability and reliability of nuclear power plant safety systems. In recognition of this, CCF data are systematically collected and analysed in the International Common-Cause Data Exchange (ICDE) Project, which was initiated in August 1994. Since April 1998, the NEA has formally operated the project. Currently eleven countries participate in the project. The ICDE collects all events where two or more identical, redundant components of a group, fulfilling the same function, have failed or were impaired due to a shared cause (ICDE events). Complete CCFs, i. e. failure of all identical, redundant components in the group due to a shared cause are an important subset of the collected data. Currently, data exchange and analysis covers the following components: centrifugal pumps, diesel generators, motor-operated valves, safety and relief valves, check valves, reactor protection system components (level measurement, control rod drives, etc), circuit breakers, and batteries. The main findings of the ICDE reports issued by 2005 show averaged over all components that about two thirds of all complete CCF events involve faulty actions by plant personnel and contractors. The single largest contribution is from faulty testing and maintenance work due to deficient and/or incomplete procedures. Other important causes are insufficient testing and requalification of components or systems after maintenance, repair, modifications or backfitting work, as well as operator errors of commission. The probability that a reported ICDE event is a complete CCF decreases strongly with increasing number of redundant components, demonstrating the effectiveness of redundancy as a powerful defence against CCFs. However, complete CCFs cannot be completely prevented by high redundancy only. (orig.)

  17. Nuclear power plant performance objectives for the 1990s

    International Nuclear Information System (INIS)

    The development of nuclear energy in recent decades as an indispensable energy source has produced a very high standard of reliability and availability. Light water reactors (PWR and BWR) constitute the basis for the commercial application of nuclear energy. Under construction at the end of 1986 were 133 units with a total capacity of c. 118,000 MW(e). Most of these units will start commercial operation between the end of the 'eighties and middle of the 'nineties. A low level increase rate of approx. 28,000 - 30,000 MW up to the year 2000 may be assumed additionally. The objectives for the 1990s are the backfitting of current LWRs as far as necessary and feasible and as a future task the development of the next generation of LWRs. The main goal is seen in cost reduction and shortening construction time to increase the economic viability of nuclear energy compared with fossil fired power plants. In the technical field the reactor core, nuclear fuel, instrumentation and control are areas for improvement. The nuclear fuel cycle will be optimized for higher burnup, lower uranium utilization by application of modified core management methods, improvements in operating flexibility and saving of natural uranium and future development of reprocessing technology and recycling of reprocessed U and Pu. Instrumentation and control systems are areas for evolutionary modifications and/or changes. With the introduction of computer systems in various fields and in the control room, the development of fibre-optical systems and the introduction of microprocessors higher reliability is possible for control of the entire plant and reduction of man-machine interfaces. The application of part task simulators and special simulators beside the well proven full scope simulator enables optimal personnel training and retraining. Another important area for future development and increased plant availability is reduction of planned and unplanned plant outages, shortening refuelling periods

  18. An analysis of nuclear power plant operating costs

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  20. Securing the Stability of Ageing Nuclear Power Plants Considering Fukushima Lessons Learned

    International Nuclear Information System (INIS)

    To avoid repeating an accident similar to the one that occurred at TEPCO’s Fukushima Daiichi plant, policy bills to reform regulatory organizations and systems were submitted to Congress. A new nuclear regulatory authority will be created by the Ministry of the Environment, aimed at obtaining a nuclear regulatory body capable of preserving total independence in its regulatory decision making processes. One of the focal points is enhancement of measures against severe accidents and reassessment of the potential effects of initiating external events. Implementation of these measures will be mandatory and defined by law. Discussions have already started to clarify the regulatory requirements to allow regulated implementation of these policies at the national level. One of the most important lessons learned is that operating plants in Japan lacked a vision for continuous safety improvement and the necessary related programmes for its implementation. The changes being proposed to enhance safety at the Japanese plants involve the basic regulatory safety framework necessary to implement any back-fitting programme and enact a comprehensive safety assessment of all the Japanese plants. The newly proposed regulatory system imposes an operation limit of 40 years to deal with aged reactors. Stress tests are being conducted to enhance public confidence in continuing nuclear power plant operation. In order to implement the safe decommissioning of the accident site in Japan, many issues are still uncertain or need to be investigated further. To that effect, and in the interest of all nuclear power operators around the world, international collaboration in R&D is strongly encouraged. The utilization of materials and components extracted from decommissioned reactors in research laboratories is strongly encouraged so as to meet the overall objective of achieving safe, LTO and effective PLiM of nuclear power plants around the world

  1. Advanced human-system interface design review guidelines

    International Nuclear Information System (INIS)

    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

  2. Methods development to evaluate the risk of upgrading to DCS: The human factor

    International Nuclear Information System (INIS)

    The NRC recognizes that a more complete technical basis for understanding and regulating advanced digital technologies in commercial nuclear power plants is needed. A concern is that the introduction of digital safety systems may have an impact on risk. There is currently no standard methodology for measuring digital system reliability. A tool currently used to evaluate NPP risk in analog systems is the probabilistic risk assessment (PRA). The use of this tool to evaluate the digital system risk was considered to be a potential methodology for determining the risk. To test this hypothesis, it was decided to perform a limited PRA on a single dominant accident sequence. However, a review of existing human reliability analysis (HRA) methods showed that they were inadequate to analyze systems utilizing digital technology. A four step process was used to adapt existing HRA methodologies to digital environments and to develop new techniques. The HRA methods were then used to analyze an NPP that had undergone a backfit to digital technology in order to determine, as a first step, whether the methods were effective. The very small-break loss of coolant accident sequence was analyzed to determine whether the upgrade to the Eagle-21 process protection system had an effect on risk. The analysis of the very small-break LOCA documented in the Sequoyah PRA was used as the basis of the analysis. The analysis of the results of the HRA showed that the mean human error probabilities for the Eagle-21 PPS were slightly less than those for the analog system it replaced. One important observation from the analysis is that the operators have increased confidence steming from the better level of control provided by the digital system. The analysis of the PRA results, which included the human error component and the Eagle-21 PPS, disclosed that the reactor protection system had a higher failure rate than the analog system, although the difference was not statistically significant

  3. 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.

  4. 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)

    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

  5. MET-RODOS: a comprehensive atmospheric dispersion module

    International Nuclear Information System (INIS)

    A comprehensive meteorological dispersion module called MET-RODOS is being developed to serve the real-time RODOS decision support system with an integrated prediction capability for airborne radioactive spread, deposition and gamma radiation exposure on all scales. Deposition, ground-level air concentrations, and ground level gamma dose rates from up to 15 simultaneous released nuclides are calculated using a nested system of local and long-range atmospheric dispersion models, driven by real-time available on-line meteorological information. The MET-RODOS module uses concurrently the available source term and weather information in the RODOS data base system, and returns, up to +36 hour forecasts of nuclei-specific air and deposited concentrations including gamma dose rate estimations for display and subsequent processing within the RODOS framework. Weather and meteorology is available to the system via on-line connections to on-site local meteorological observations (met-towers and sodars) and via network (either public-domain Internet or user-owned point-to-point ISDN) connections to remote national or international meteorological forecasting services. In its final form, scheduled for operational use in 1999, the MET-RODOS meteorological module is intended to service the RODOS system with actual and forecast (+36 hour) nuclei-specific air concentrations, deposition values, and gamma radiation estimates on the local, national, and European scale. Provisions are furthermore being made for accommodating on-line available radiological monitoring data in the meteorological model chains in order for the module to assist with source term determination based on real-time data-assimilation and back-fitting procedures. (orig.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  8. Advances in nuclear fuel technology. 2. Advances in nuclear fuel technology for LWRs

    International Nuclear Information System (INIS)

    From a viewpoint of upgrading on economical efficiency, some developments aiming at reduction of fuel cycle cost and used fuel forming amounts and response to long term operation cycle are carried out. These developments are required for back-fitness for already established reactors, so are progressed under limited changing tolerances. On fuels for BWRs, under confirming their used results, stepwise planning on upgrading of burnup such as steps 1, 2 and 3 is examined. For example, on a new 8 x 8 zirconium liner fuel (step 1), by adapting a zirconium liner cladding tube, its PCI (fuel pellet-cladding interaction) resistance feature is largely improved, to reach about 33 GWd/t in average discharge burnup. And, a high burnup 8 x 8 fuel (step 2) is intended to upgrade high burnup by increasing concentration degree as well as to improve design on fuel assembly structural element, to further upgrade its economical efficiency. At present, on a 9 x 9 type fuel (type 3) begun on its practical use, array of fuel rods is made by nine rows and nine columns, to increase to 45 GWd/t in average discharge burnup and 55 GWd/t in highest assembly burnup. Furthermore, on future fuel, a wide high burnup over limitation on improved 9 x 9 type and fuel cycle is investigated, to promote developments on improved fuel pellet and new alloys for structural materials. Here were introduced design and production based on upgradings of reliability and economical efficiency on recent commercial LWRs, and trends on their R and D at every fields. (G.K.)

  9. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

    In beginning, a history of legislative process regulating industrial utilisation of nuclear energy is given, including detailed list of decrees issued by the first regulatory body supervising Czech nuclear installations - Czechoslovak Atomic Energy Commission (CSKAE). Current status of nuclear regulations and radiation protection, especially in connection with Atomic Act (Act No 18/1997 Coll.), is described. The Atomic Act transfers into the Czech legal system a number of obligations following from the Vienna Convention on Civil Liability for Nuclear Damage and Joint Protocol relating to the Application of the Vienna and Paris Convention, to which the Czech Republic had acceded. Actual duties and competence of current nuclear regulatory body - State Office for Nuclear Safety (SUJB) - are given in detail. Execution of the State supervision of peaceful utilisation of nuclear energy and ionising radiation is laid out in several articles of the Act, which comprises: control activities of the SUJB, remedial measures, penalties. Material and human resources are sufficient for fulfilment of the basic functions for which SUJB is authorised by the law. For 1998, the SUJB allotted staff of 149, approximately 2/3 of that number are nuclear safety and radiation protection inspectors. The SUJB budget for 1998 is approximately 180 million Czech crowns (roughly 6 million US dollars). Inspection activity of SUJB is carried out in three different ways: routine inspections, planned specialised inspections, inspections as a response to a certain situation (ad-hoc inspections). Approach to the licensing of major plant upgrades and backfittings are mainly illustrated on the Temelin NPP licensing. Regulatory position and practices concerning review activities are presented. (author)

  10. MOX fuel design and development consideration

    International Nuclear Information System (INIS)

    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 UO2 fuel, because Pu content of MOX fuel for Pu thermal utilization is low. The fuel design code for UO2 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 UO2 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 UO2 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

  11. Evaluation of post-accident combustible gas control at Millstone Unit I

    International Nuclear Information System (INIS)

    On December 2, 1981 the NRC published 10CFR 50.44, Interim Requirements Related to Hydrogen Control. The major impact of this regulation on Mark I BWR Containments was to require inerting and to eliminate containment purging as a post-accident combustible gas control scheme. This rule in effect required installation of hydrogen recombiners on all plants originally licensed assuming purge/pressurization for hydrogen control. Because of the high costs of backfitting hydrogen recombiners on Millstone-I, it was decided to more thoroughly investigate the technical bases of the need to purge containment for hydrogen control if the containment was initially inerted. Northeast Utilities Service Co. conducted this investigation and identified the following: a) preinerting of a Mark I Containment with excess N2 is so effective that no extent of metal-water reaction can directly lead to flammable H2-O2 gas mixtures; b) production of hydrogen via large metal-water reactions actually makes the containment more inert by further diluting the initial O2 content; c) with removal of oxygen (air) in-leakage to the containment from the M.S.I.V. and safety relief valve control air lines, radiolysis is the only credible oxygen source present in the post-accident containment environment; d) post-accident boiling phase radiolysis, which is the most rapid source of oxygen generation, is insufficient to yield a flammable mixture; e) post-accident subcooled phase radiolysis in which natural recombination is strongly favored, yields stable equilibrium levels of H2 and O2 below the flammable mixture level

  12. Improved once-through fuel cycles for light water reactors

    International Nuclear Information System (INIS)

    This paper is being presented at this time to provide preliminary technical and economic data to INFCE for use in comparisons of alternate nuclear systems. Programs to develop improved once-through fuel cycles for the light water reactor are under way in the United States; therefore, the information presented in this report is preliminary and will be updated in the future as it becomes available. In the meantime, the following limitations should be recognized when using the information in this report: 1. The paper quantifies fuel utilization improvements which should be technically feasible in reactors now operating or under construction and indicates the approximate time frame when the necessary development and demonstration could be completed. It does not attempt to estimate the rate at which these improvements would attain acceptance and use by the industry. 2. One particular set of PWR and one particular set of BWR nuclear reactor and fuel design characteristics are used as base cases, from which many of the improvements are estimated. Many plants operating and being built throughout the world of course differ in design features, fuel management schemes, and fuel utilization efficiencies from the base cases used in this paper. The degree of improvement obtainable in these other designs, for each type of change considered, will vary with each design. 3. The changes emphasized here could all be backfitted in existing plants. Other possible improvements are limited by the need to avoid reducing the power output or capacity factor of the plants. New plants could be designed to accommodate such changes without reducing the power output or capacity factor. This could yield greater improvement in fuel utilization than can be obtained in existing plants. This longer range potential has not been examined here

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

    International Nuclear Information System (INIS)

    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

  14. EU stress test: Swiss national report. ENSI review of the operators' reports

    International Nuclear Information System (INIS)

    The earthquake on 11 March 2011 and the resultant tsunami led to severe accidents with core melt in three nuclear power plants (NPP) units at the Fukushima Dai-ichi site. These events were classified by the Japanese authorities as 'major accident' (INES 7). The EU stress test is part of the review process which Switzerland initiated immediately after the reactor accident. The Swiss Nuclear Safety Authority (ENSI) required from the operators of the Swiss NPPs to implement immediate measures and to conduct additional re-assessments. 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 re-assessments, which were to be carried out immediately, focused on the design of the Swiss NPPs against earthquakes, external flooding and a combination thereof, as well as investigations on the coolant supply for the safety systems and the spent fuel pool cooling. ENSI carried out an analysis of the events at Fukushima and published the results providing detailed descriptions of the causes, consequences and radiological impacts of the accident. The purpose of the EU stress test is to examine the robustness of the NPPs in case of impacts beyond the design basis due to earthquakes, external flooding and extreme weather conditions, loss of power supply and heat sink, and severe accident management. As the first step, it was necessary to present the hazard assumptions and design bases for the NPPs, and to assess their adequacy. In the second step, the objective was to identify and evaluate the protective measures implemented and their safety margins as compared to the design. Improvement measures were to be derived. The review by ENSI confirmed that the Swiss NPPs display a very high level of protection against the impacts of earthquakes, flooding and other natural hazards, as well as loss

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

    International Nuclear Information System (INIS)

    reasons as TEPCO in Japan until Fukushima. The EU tried to respond to this ''new experience'' of Fukushima by conducting the stress tests and hoping that the results will lead to higher safety. This report investigated the result, the very concrete measures each nuclear safety authority will require its operators to implement and until which date. Transparency is another important tool to control nuclear risk; while ENSREG certainly recognizes this fact, not all national nuclear regulators and operators act accordingly to fulfil this need of higher transparency. It is evident that some countries treated this task rather as a formality or paperwork than a plant safety upgrade program. (The ENSREG peer review hopefully will insist on introducing additional measures to the national plans in those cases where the national regulator required less safety measures than the stress tests peer review recommended.) In general, there are different possibilities for operator and nuclear authority to remedy the shortcomings the stress tests revealed: - A quick response, but without any guarantee that the measures are sufficient (e.g. Wylfa, UK). - A comprehensive evaluation of possible hazards and protections measures, which will take more than ten years (e.g. Gravelines and Cattenom, France). - Business-as-usual, (e.g. Temelin, Czech Republic). The idea of the stress tests is more or less ignored. Instead the already ongoing measures are listed, major hardware improvement avoided. None of those possible variants increase the nuclear safety to an acceptable level. The very obvious solution - permanent shut down - needs to be considered and is in several cases the only safe option. This applies in particular to those plants where significant improvements cannot be achieved by the planned deployment of mobile equipment only or by having plants on the grid in the current status for many more years while evaluations and assessment are under preparation and again later backfittings would

  16. 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 take over ten years time. The National Reports are heavily relying on the new magic solution to severe deficiencies at the plants due to design or the site: mobile equipment, which is easy to plan and store in the plant and therefore a cheaper solution compared to comprehensive measures. But under severe accident conditions, it is very unlikely that the proposed mobile equipment can be put to work as quickly as necessary; to rely to such a large extent on manual actions is in regard of the consequences of a severe accident irresponsible. Furthermore, the new mobile equipment is useless if the staff training and response during the accident is not perfectly according to plan. However not only the ''know-how'' but also the ''know-why'' is very important. This is also one important lesson learnt from the Fukushima accident. Limited backfitting measures do not significantly improve the safety level because they cannot compensate the increasing threat of hazards (e.g. by climate change) and of ageing effects. Furthermore, the experiences show that back-fitting measures could cause new faults (e.g. because of defective mounting, forgotten scrap etc.). Comprehensive plant modifications which would actually improve the safety level are technically impossible or would be done only in exchange for prolonged operation times, at the same time carrying the risks of aging plants as mentioned above.

  17. EU stress test: Swiss national report. ENSI review of the operators' reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-12-15

    The earthquake on 11 March 2011 and the resultant tsunami led to severe accidents with core melt in three nuclear power plants (NPP) units at the Fukushima Dai-ichi site. These events were classified by the Japanese authorities as 'major accident' (INES 7). The EU stress test is part of the review process which Switzerland initiated immediately after the reactor accident. The Swiss Nuclear Safety Authority (ENSI) required from the operators of the Swiss NPPs to implement immediate measures and to conduct additional re-assessments. 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 re-assessments, which were to be carried out immediately, focused on the design of the Swiss NPPs against earthquakes, external flooding and a combination thereof, as well as investigations on the coolant supply for the safety systems and the spent fuel pool cooling. ENSI carried out an analysis of the events at Fukushima and published the results providing detailed descriptions of the causes, consequences and radiological impacts of the accident. The purpose of the EU stress test is to examine the robustness of the NPPs in case of impacts beyond the design basis due to earthquakes, external flooding and extreme weather conditions, loss of power supply and heat sink, and severe accident management. As the first step, it was necessary to present the hazard assumptions and design bases for the NPPs, and to assess their adequacy. In the second step, the objective was to identify and evaluate the protective measures implemented and their safety margins as compared to the design. Improvement measures were to be derived. The review by ENSI confirmed that the Swiss NPPs display a very high level of protection against the impacts of earthquakes, flooding and other natural hazards, as

  18. Characteristics and use of the component event data bank

    International Nuclear Information System (INIS)

    The Component Event Data Bank (CEDB) is a centralized bank collecting, at the European level, data describing the operational behaviour of components of Nuclear Power Plants (NPP's) operating in various European countries. As of December 1987, the CEDB contains data from about 5,200 mechanical and electromechanical components, pertaining to 21 component-families and 51 engineering systems, monitored for an averaged time of 5 years in 10 LWR Units (and in the steam-water cycle of other reactor type Units or conventional Units) located in 7 European countries; the failure-events recorded are about 4,200. The CEDB has as main objective the promotion of safety. In particular it has been conceived as a support to the analyst in his safety assessment for the design of a new NPP or the backfitting of an old one. By putting together the operational experience of European NPP's, it was the intention to create a database of raw data, to be conveniently processed, in order to: improve the credibility of existing estimated reliability parameters by exploiting the necessary feed-back from plant operation; provide a solution to one of the major problems of the reliability analyst; namely the wide spread in reliability parameters existing in the current literature, especially for mechanical and electromechanical components; allow comparison between the performance of components of plants of different countries. This paper shortly describes: the main features of the bank classification scheme, its data retrieval capabilities and on-line statistical processing programmes; some improvements under study to better meet PSA needs. Some examples of on-line data treatment are given and commented on. The structure of an organized output (a CEDB-Reliability Data Book), which is being implemented, is shortly illustrated. The results of a study on linked multiple failure-events and some analyses based on the application of multivariate analysis techniques are summarized and the interest to

  19. 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

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

    International Nuclear Information System (INIS)

    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

  1. Analysis and prognosis of radiation exposure following the accident at the Siberian chemical combine Tomsk-7

    International Nuclear Information System (INIS)

    -line puff diffusion model, in backfitting mode. (EG) (13 tabs., 20 ills., 15 refs.)

  2. Safeguards-by-design: 3S integration

    International Nuclear Information System (INIS)

    Design and construction of a nuclear energy system, particularly the first of its kind - or the first for a given host state - is an immensely complex undertaking. Additionally safeguards, security, and safety each have significant requirements that must be properly addressed by the facility and system design effort. However, the requirements are often in apparent conflict. A simple example would be the safety requirement for emergency egress doors contrasted against the physical security requirement to limit and control access points. This conflict is significant, even as it is often understated. The potential for conflict is increased when different parties are responsible for each component, as in the International Atomic Energy Agency (IAEA) performing safeguards while the host state provides for safety and security. Early and complete identification of requirements is critical to design success. This allows consideration of intrinsic design features that can satisfy all parties. The majority of the cost to design and build a facility is committed by the end of the conceptual design. As the design progresses, the cost to change the facility via retrofits is increasingly, even prohibitively, expensive, both in money and time. Conflicts among the safeguards, security, and safety design requirements must be resolved early to allow optimal design solutions, and to avoid costly backfits. The Safeguards-by-Design (SBD) project currently underway in the United States is developing a design process that will fully integrate safeguards (to include state and international safeguards, and proliferation barriers) and security into the facility design process, with close coordination with safety. The SBD process includes key features such as early safeguards and security input to the facility requirements documents, a systems approach to nonproliferation and security considerations within the design effort, and agreed upon timelines for communicating with the IAEA. The

  3. ACR-1000: Operator - based development

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU* reactors to establish Generation III+ Advanced CANDU ReactorTM (ACRTM) technology. The ACR-1000TM 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 CANDUTM 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 main program

  4. Hardware replacements and software tools for digital control computers

    International Nuclear Information System (INIS)

    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

  5. 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)

    Atomic energy law remains a living, fascinating subject matter. Nearly 200 participants were convinced of this impression at the 14th 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 14th 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 14th 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, material

  6. Level-2 Probabilistic Safety Assessment for 220 MWe Indian PHWR (KAPS)

    International Nuclear Information System (INIS)

    Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used to complement, enhance and validate conclusions that are based on deterministic design principles. This paper discusses various stages and insights drawn from Level-2 PSA study performed for Kakrapara Atomic Power Station (KAPS), an Indian PHWR. The Level-2 PSA deals with frequency and magnitude of releases to environment and consists of probabilistic and deterministic analysis elements. The probabilistic element consists of the development and quantification of containment logic models for each PDS. The deterministic element consists of calculating the release magnitude from the core, physical process of accident progression including containment response and source term analysis of radionuclide releases to the environment for the representative events from each PDS using appropriate codes. Some intended objectives of this Level-2 PSA study were : - To gain insights into the progression of severe accidents and containment performance and identify and prioritise scenarios requiring further and more refined analysis. - To identify major containment failure modes and to estimate the corresponding releases of radionuclides. - To identify any weak links in the plant design and suggest plant specific back-fit measures as risk reduction options. The Level-2 is an extension of Level-1 PSA The Plant Damage States form the interface between the two analyses are developed according to the requirement of Level-2 analysis. The containment Engineered Safety Features (ESFs) are treated as part of the Level-2 analysis. The ultimate product of a Level 2 PSA, is a discussion of a number of challenges to the containment, of the possible containment responses and their estimated probabilities and an assessment of the consequent releases

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

    International Nuclear Information System (INIS)

    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)

  9. Safety performance indicators

    International Nuclear Information System (INIS)

    Since its creation the nuclear industry has been struggling with the question of how safe is safe enough. Safety is a goal common to all involved in the design, operation and regulation of a nuclear installation. As a concept, safety is not easy to define. However, there is a general understanding of what attributes a nuclear power plant should have in order to operate safely. The challenge lies in measuring the attributes. After accumulating more than forty years of experience, through learning and safety backfitting, the nuclear industry and regulators are ready to use that operating experience and risk analysis to focus on the most significant aspects of plant safety. Nuclear power plants are developing human performance, self-assessment and corrective action programmes to improve safety and production. Operating experience is shared throughout the nuclear industry and incorporated into these programmes. Probabilistic safety assessment techniques have also improved and expanded so that risk insights are being increasingly used. In the regulatory field there is a clear tendency to combine the traditional deterministic and the risk informed, performance based ways of looking at what is important to ensure public health and safety. Risk informed, performance based regulation needs a comprehensive set of safety indicators as an important oversight tool. At the same time it is necessary to preserve the current regulatory requirements and criteria especially in safety areas, which are not covered by the set of indicators. A new approach should be designed to fulfill the needs of nuclear power plants while maintaining an effective regulatory oversight programme. There are two main driving forces to use safety Performance Indicators. The first requires nuclear power plants to collect data and process a set of indicators prescribed by the regulatory body, or recommended by national or international organizations. The second is a voluntary approach which depends on

  10. Status of subjects related to lifetime extension of nuclear power plants in Argentina

    International Nuclear Information System (INIS)

    The possibility of lifetime extension in nuclear power plants depends on maintenance, safety assessment, corrective actions and continuous improvement during plants life. Atucha I nuclear power plant has a backfitting program ongoing, all the coolant channels have been replaced, the Emergency Power Supply System has been modified, a Second Heat Sink is under construction, and all the tube guide control rods were replaced, as well as the In core Neutron Flux Sensors guide tubes. Moreover, several safety assessments have been made, among which a plant specific Probabilistic Safety Assessment, which includes fire risk analysis, addressing radioactive releases different from reactor core ones, and a PSA during shutdown. Last of all, an analysis of the introduction of slightly enriched fuel elements into the core, and a pressure vessel integrity analysis were made. With respect to the pressure vessel integrity analysis, the Responsible Organization is working with the purpose of reducing the uncertainties associated to thermohydraulic, neutronic and fractomechanics calculus, looking for higher safety margins that will allow them to ask the Regulatory Body for a life extension. At Embalse nuclear power plant there exists an In Service Inspection Program to examine and verify the state of components and/or systems, to enable the prediction of any eventual damage and determine the corresponding corrective actions to preserve safety functions. The Responsible Organization also created the Technical Committee of Channels Subprogram for CNE to co-ordinate the task related to the pressure tube inspection program. A set of (ultrasonic and eddy current) measurement techniques of the pertinent parameters, as well as the 'Spacer Location and Repositioning' program, has been applied in CNE since the late eighties. The Regulatory Body required the Responsible Organization the elaboration of a Level 1 probabilistic safety assessment, aiming at evaluating CNE safety level, in order to

  11. Cost effective decommissioning and dismantling of nuclear power plants

    International Nuclear Information System (INIS)

    lessons learned from previous experience and on consistent application of methods and processes applied in new builds and large projects for modernization and back-fitting of existing plants. (orig.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  14. Hot functional testing of the pressurized heavy water reactor plant Atucha II with light water

    International Nuclear Information System (INIS)

    The two pressurized heavy water reactors PHWR Atucha I (designed and built by S/KWU, now AREVA), and Atucha II (designed by S/KWU and plant construction now completed by NA-SA) are owned by Nucleoelectrica Argentina S.A. (NA-SA). Atucha II was designed in the 1980'ies in parallel to the two most recent S/KWU PWR generations Prekonvoi and Konvoi. Its basic design has been updated and optimized including also backfitting of components and systems for severe accident management. The gross electric power of the plant is 745 MWe. Construction and commissioning of Atucha II has been resumed by NA-SA after a work stop in the 1990'ies and is now almost completed. Hot functional testing HFT was performed in two phases in September and October 2013 and in March and April 2014. Hot functional testing was performed with light water and the fuel assemblies loaded. The chemistry program for the HFT was derived from practices and experience gathered at other S/KWU designed PWRs during HFTs and consisted of the following main targets and requirements: (1) Low chloride and sulfate concentrations close to normal operation values specified in the VGB water chemistry guideline for power operation of PWR plants; (2) Thorough oxygen removal during heat-up and reducing conditions through N2H4 dosing; (3) High pH value (target range 1.5 to 2 ppm Li); (4) Passivation treatment of the nuclear steam supply system NSSS at temperatures of at least 260°C for a time period of at least 120 hours; (5) Zinc addition at a constant rate of 20 g Zn per day throughout the various HFT phases. Zinc dosing was begun during the first heat-up of the plant at temperatures above approx. 150°C. Daily measurement of the zinc concentration for process control was not necessary and not required due to the elaborated zinc application procedure. The main results of the chemistry program for the HFT of plant are described and evaluated in this contribution. Data shows that all chemistry targets were met

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

    International Nuclear Information System (INIS)

    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

  16. Nuclear power plant life management: Strategy for long term operation of the Beznau NPP unit 1 and 2

    International Nuclear Information System (INIS)

    Full text: Both Beznau nuclear power plants (KKB) are of the Westinghouse two loop PWR type with a rated electrical output of 380 MWel and have been operated on base load demand since 1969 and 1972 respectively. The design base uses rather conservative assumptions. By performing selective measures, such as large investments in backfitting, the safety of the plants has been continuously enhanced. Through focussed modernisations and careful maintenance, the overall condition of both Units is excellent. A comprehensive ageing management programme (AMP) for all safety related structures, systems and components (SSC) was set up and started 15 years ago, according to the requirements of the Swiss Federal Nuclear Safety Inspectorate (HSK). The AMP contains all essential actions of evaluation and control of the ageing concerning the material and also conceptual aging. The permanently increasing demand for electrical energy in Switzerland, as well as economic factors, led NOK to evaluate long term operation (LTO) of KKB. Beside hydro power, nuclear power covers 40% of the demand in Switzerland. A new nuclear law in Switzerland allows unlimited operation of a NPP as long as safety goals are met. The performed analysis and conclusions for a potential LTO is presented in the paper, including aspects such as technical issues, fuel, radwaste, elusive risks, personnel management and economics to assure LTO with the best achievable safety level. The necessary actions to address and control the ageing mechanisms of the civil structures have been established in line with the AMP. The procedures for implementing these actions are in place and are continuously executed. The technical feasibility for the essential actions required for a LTO up to 60 years is given. A systematic registration of the actions to control ageing of all electrical systems and components occurs in line with the AMP. The main challenge is not the material related ageing but the availability of products and

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

    International Nuclear Information System (INIS)

    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. ILK statement on determining operation periods for nuclear power plants in Germany; ILK-Stellungnahme zur Festlegung von Betriebszeiten fuer Kernkfraftwerke in Deutschland

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-09-15

    period of 40 years at the latest, a special safety review should be supplied by the licensee and be evaluated by the authority. On this basis plant operation can continue for 10 more years at a time as insofar as the authority does not raise objections. In addition to the analyses covered by the PSR, the special safety review contains the following requirements: - The current status of the plant or its status at the start of the renewal period is to be compared to the requirements of the safety criteria and the RSK safety standards. - Operating management is carried out according to the best current practices. - An effective aging management exists. - An up-to-date probabilistic safety analysis (PSA) that covers all operating conditions exists for Level 1 and Level 2. - Backfits that are necessary for maintaining the existing safety level or lead to a further improvement of the safety level when taking the appropriateness of means into account have been or will be applied. (orig.)

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

    International Nuclear Information System (INIS)

    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

  20. 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

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

    International Nuclear Information System (INIS)

    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

  2. Fundamental study on serious accidents and their management in fuel fabrication/enrichment facilities and reprocessing facilities

    International Nuclear Information System (INIS)

    The 'Act for the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors' was amended and issued in June 2012 taking into account the lessons derived from the accident of TEPCO Fukushima Daiichi Nuclear Power Plant occurred in March 2011. The main amendments were as follows; Preparation for the management of serious accidents, Introduction of evaluation system for safety improvement, Application of new standards to existing nuclear facility (back-fitting). Japan Nuclear Energy Safety organization (JNES) started this fundamental study on serious accidents and their management, as a safety studying in fuel fabrication/enrichment facilities and reprocessing facilities, for the purpose to contribute to the implementation of new Rules by Nuclear Regulation Authority. From the technical view to be concerned such as fundamental concept of the Rules and applicability of risk-informed regulation, the following 7 subjects were studied: 1) Application concept of the defense in depth to these facilities. 2) Positioning of serious accidents and their management in the defense in depth. 3) Definition of the serious accidents in these facilities. 4) Postulated external events for the study of the serious accidents and their management. 5) Objectives and requirements of the accident management (assurance of reliability). 6) Confirmation logic flow on sequence of the serious accidents and the accident management measures. 7) Applicability of risk information. During the study on these subjects, features of the facilities were clarified at first. Based on concept of the defense in depth, which is the basic principle in safety, and referring to information related to domestic/foreign serious accidents, JNES conducted the fundamental study and made the following suggestions: 1) Definition of the serious accidents of the facilities. The definition is expected to contribute the discussion on new Rules by Nuclear Regulation Authority. 2) Methodology to examine the

  3. 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)

  4. Aging evaluation methodology of periodic safety review in Korea

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  6. EC Research Contribution to Decision-making Processes Relevant to Severe Accident Management

    International Nuclear Information System (INIS)

    As a result of the two well-known civil nuclear accidents and of the consequent increase in safety requirements, the need to properly assess severe accident (SA) scenarios for present and future nuclear power plants (going beyond the traditional three-level defence-in-depth strategy) became evident. In this line, various research activities were launched and are performed within the Euratom Framework Programmes, in particular the completed Fourth one (F P-4, 1994-1998) and the present Fifth one (FP-5, 1998-2002). The initial orientation of the EC research activities was mainly focused on improving the understanding of the phenomena and mechanisms involved in such accidents, in order to contribute to prevent possible final radioactivity releases. A consensus on how to model those SA phenomena in accident safety analyses by means of specific tools (SA codes developed, verified and validated through experimental results provided) is reasonably advanced. Currently, the EC research activities related to severe accidents are balanced between a twofold approach aimed at assessing the risks related with severe accident scenarios and to support the development of severe accident management (SAM) strategies, together with the optimisation of backfitting measures for existing reactors or specific designs for future nuclear power plants. This new orientation is confronting difficulties, inherent to the phenomenological character of several research activities, which make a direct application of the results into SAM measures premature in some cases. In this regard, this paper presents a series of ten selected FP-5 projects with emphasis placed on the applicability of research results towards SAM strategies to be used by decision-makers amongst utilities, the nuclear industry in particular designers, and regulators. The majority of them also contain -further to the SAM approach- supporting elements focused on risk assessment. The revised programme of the key action 'Nuclear

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

    International Nuclear Information System (INIS)

    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 5th 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 4th 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 1st 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 in 2005

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

    International Nuclear Information System (INIS)

    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

  9. Essential severe accident mitigation measures for operating and future PWR's

    International Nuclear Information System (INIS)

    Severe Accident mitigation measures are a constituent of the safety concept in Europe not only for operating but also for future light water reactors. While operating reactors mainly have been backfitted with such measure, for future reactors Severe Accident mitigation measures already have to be considered in the design phase. Severe Accident measures are considered as the 4th level of defense for future reactors. This difference has consequences also on the kind of measures proposed to be introduced. While in operating plants Severe Accident mitigation measures are considered for further risk reduction, in future reactors an explicit higher level of safety is required resulting in additional design measures. This higher safety level is expressed in the requirement that there must be no need for evacuation of surrounding populations except in the immediate vicinity of the plant and for long-term restrictions with regard to the consumption of locally grown food. Because of the potential hazard posed by radioactive releases to the environment in the event of an Severe Accident situation depends largely on the airborne material in the containment atmosphere and on the containment integrity, new system features to prevent loss of containment integrity have been introduced in the design of the NPP's. For these tasks it has been necessary to develop and qualify new system technologies and implement them finally into NPP's, e.g. like systems for containment atmosphere H2-control, filtered venting, core retention devices and atmosphere sampling. The following systems are introduced for operating as well as for future plants: · The Hydrogen Control System is based on the Passive Autocatalytic Recombiner (PAR) technology. There is no need for any operator actions because of the self-starting feature of the catalyst if hydrogen is released. · In situ Post Accident Sampling System (In situ-PASS) are introduced for the purpose of obtaining realistic information on airborne

  10. Proceedings of the OECD/NEA workshop on seismic risk - Summary and conclusions - Committee on the Safety of Nuclear Installations PWG3 and PWG5

    International Nuclear Information System (INIS)

    The objectives of the Workshop were: - To provide a forum to review the recent advances in methodology and application of seismic probabilistic safety assessment and seismic margin analysis of nuclear installations, - To discuss the effective uses of the seismic PSA/margin analysis with consideration of merits and limitations of probabilistic methods, - To review the state of the art methodology to provide guidance for conducting seismic PSA, and - To discuss methodological issues and identify areas in which further research is needed for enhancing the usefulness of seismic PSA. The emphasis of the Workshop was placed on the exchange of ideas on effective ways of using seismic PSA rather than the numerical PSA results for specific plants such as core damage frequencies or seismic hazard. From the presentations and discussions in this workshop, it can be concluded that the seismic PSA/Margin methods have been and are being used world-wide, providing useful information for safety improvement or decision making, and great amount of experience has been accumulated, although the status of programs in member countries vary widely. The objectives of such studies include the following: - To examine whether there are cost effective ways to improve safety from ALARP point of view - To assist in decision making in backfitting by identifying cost effective improvements - To demonstrate the seismic margin of existing or future plants - To examine the vulnerabilities in protection against severe accident - To improve design of future reactors by identifying relatively weak points - To assist in selection of new sites for NPPs. Although numerical results from seismic PSA have not been directly used in seismic design as an alternate or supplement to current deterministic analysis methods, some countries have already adopted the use of probabilistic seismic hazard analysis for determining design basis earthquakes (SSE in USA) and some activities are ongoing to develop methods for

  11. 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

  12. 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

  13. Containment In-Situ PASS and Emission Monitoring System Design and Implementation

    International Nuclear Information System (INIS)

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Europe have been back-fitted with supplementary systems such as containment venting systems and hydrogen control systems. The potential hazard posed by radioactive releases to the environment in the event of an accident in a nuclear power plant largely depends on the airborne material in the containment atmosphere, like aerosol-bound and non aerosol-bound radionuclides. In conjunction with this the German Reactor Safety Commission (RSK) imposed the additional recommendations that provisions be made for: - Post accident sampling of the containment atmosphere for the purpose of obtaining information on the condition of the core and on potential hazards to the environment and - Emission monitoring for containment venting. Containment In-situ PASS: Various systems for use in German PWRs and BWRs have been investigated for their capability to take a representative probe of the radioactive nuclides in the containment atmosphere. Investigations showed that systems already installed in power plants - e.g. with sampling line lengths of 30 to 50 m and extraction pipe systems equipped with ball valves for sampling, exhibit considerable problems due to deposition inside the pipes. Deposition rates determined from various experiments resulted, when extrapolated for iodine and aerosols in sampling systems with sampling line lengths of 30 to 80 m, in deposition loss factors of 10 to 100 and more. As these problems cannot be solved satisfactorily with conventional sample piping systems, a new in-situ sampling technique has been developed together with the German utilities and the first full-scale prototype In-Situ PASS (Post-Accident Sampling System) for Germany's nuclear power plant unit Neckarwestheiml has been tested under all relevant containment conditions. The thermal-hydraulic performance as well as retention and deposition properties of this equipment were

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

    International Nuclear Information System (INIS)

    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

  15. Legal and regulatory aspects of long-term operation of nuclear power plants in OECD member countries

    International Nuclear Information System (INIS)

    progress in science and technology, feedback from operating experience and lessons learnt after major nuclear accidents, such as Three Mile Island, Chernobyl or, most recently, Fukushima. Within this framework, economic reasons will usually drive long-term operation of nuclear reactors. Operators will seek permission to continue operating a nuclear power plant beyond 30 or 40 years if this is economically viable given the investments necessary to continue to comply with the nuclear safety framework as required by regulators. On the one hand, substantial backfitting may be required due to regulatory requirements. On the other hand, the capital costs to allow long-term operation of nuclear power plants may be much smaller than investment in any type of replacement capacity. In many OECD member countries utilities tend to choose continued operation of existing nuclear reactors as the cheaper and less risky alternative to new build. Indeed, there are many hurdles which new builds have to overcome, for example, unstable financial markets, complicated and unpredictable licensing procedures, public opposition to nuclear, lost experience from earlier construction and general shortage of skills. Governments may equally prefer long-term operation of nuclear reactors because it allows their countries to continue benefiting from a diversified energy mix and to enhance security of supply. Environmental considerations may constitute additional justification for favouring continued operation of nuclear reactors. Indeed nuclear plants are carbon free, as opposed to gas and coal installations, allowing governments to meet their greenhouse gas emission reduction targets. However, governments and regulators will only agree to long-term operation as long as all systems, structures and components of the installation continue to function as determined by the licence. It is therefore essential to understand the role of the licence when analysing the operation of nuclear reactors beyond the time

  16. Topical issues in nuclear safety: Proceedings of an international conference

    International Nuclear Information System (INIS)

    In 1991, the IAEA organized an international conference entitled 'The Safety of Nuclear Power: Strategy for the Future'. Recommendations from that conference prompted actions in subsequent years to advance nuclear safety worldwide. One of those actions was the establishment of the Convention on Nuclear Safety, which entered into force in October 1996. In 1998, the Agency held a conference on 'Topical Issues in Nuclear, Radiation and Radioactive Waste Safety'. The nuclear safety issues discussed were: (i) safety management; (ii) regulatory strategies; and (iii) backfitting, upgrading and modernization of nuclear power plants. Senior nuclear safety decision makers at the technical policy level reviewed these issues and formulated recommendations for future actions by national and/or international organizations. On the safety management issue, recommendations were made to monitor safety performance by using indicators. Recommendations on the regulatory strategies issue indicated the need for further work on utilizing probabilistic safety assessment and on optimizing the prescriptive nature of regulations, as well as on the future availability of competent professionals. Substantial progress has been made, and continues to be made, by Member States in enhancing the safety of nuclear power plants. At the same time, the safety standards for research reactors are being updated and new standards are planned on the safety of other facilities in the nuclear fuel cycle. In the light of these developments, it was considered appropriate to convene another conference on the following current topical issues: Risk informed decision making, Influence of external factors on safety, Safety of fuel cycle facilities, Safety of research reactors, Safety performance indicators. The conference had the objective of fostering the exchange of information on these topical issues in order to consolidate an international consensus on these issues, on the priorities for future work and the on

  17. 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)

    , 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)

  18. ACR-1000 - Designed for constructability

    International Nuclear Information System (INIS)

    Full text: One of the key aspects to be considered in the delivery of a Nuclear Power Plant is the security of the construction schedule and the need for lower construction costs. Many industries are using skids, modules and prefabrications to enhance construction productivity, reduce schedules and thus reduce costs. The leaders in this regard have traditionally been in the off-shore oil and gas, chemical, refinery and ship building industries. The concept of using modules has been utilized in Nuclear Power Plant design and construction. Atomic Energy of Canada Limited (AECL) has had considerable success at the Qinshan Nuclear Power project in China with the use of modularization, which proved extremely effective in the ability to organize parallel construction activities and shortening the schedule. Extensive use has been made of skids and modules in Japan and this also has proven effective in shortening schedules in the construction of nuclear power plants. Secondary benefits of modularization and prefabrication include decreased site congestion and logistical issues, increased worker safety and better quality control of fabrication. Modules and prefabrication allow work to be shifted to areas where skilled trades are more readily available from a site where skilled trades are very limited. One of the objectives of the ACR-1000 project is to produce a design that allows for a very secure construction schedule. The construction method and strategy, consisting of extensive use of prefabrication and modularization was defined very early in the ACR-1000 conceptual phase of the layout and design process. This has been achieved through a constructability programme that integrates the civil design with site erection and module installation. This approach takes the concept of modularization to an entirely new level, in which the use of modules is built into the design from the start, rather than backfitting modular construction into a conventionally designed plant. This