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 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)

  4. 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.)

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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)

  10. 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

  11. 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

  12. Experience in backfitting of the water treatment system in the waste-to-energy plant at Essen-Karnap; Erfahrungen bei der Umruestung der Wasseraufbereitungsanlage im MHKW Essen-Karnap

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, M.E.; Badewitz, T. [RWE Power AG, Essen und Koeln (Germany)

    2004-07-01

    RWE Power AG operates a waste incineration plant in Essen-Karnap that incinerates up to 745.000 t domestic waste and domestic waste like industrial refuse. Raw water for plant operation is taken from the Rhein-Herne canal and is processed for process water, cooling water and boiler feedwater in three steps. As of 31st May 2002, discharge of gravel filter backwash containing more than 50 mg/l suspended solids was not allowed anymore. These altered conditions of water laws required a modification of the water treatment plant. (orig.)

  13. 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

  14. 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)

  15. 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.)

  16. 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...

  17. 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

  18. 77 FR 39899 - Technical Corrections

    Science.gov (United States)

    2012-07-06

    ...), the compound word ``defense in depth'' was not hyphenated. In these paragraphs, the compound word ``defense in depth'' is revised to read ``defense-in-depth'' for consistency with Sec. 73.55(b)(3)(ii). In... Writing,'' published June 10, 1998 (63 FR 31883). VII. Backfit Analysis The NRC has determined that...

  19. 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)

  20. 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...

  1. 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.)

  2. 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.

  3. 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.

  4. 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...

  5. 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.)

  6. 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

  7. 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)

  8. 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

  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. 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

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. On concurvity in nonlinear and nonparametric regression models

    Directory of Open Access Journals (Sweden)

    Sonia Amodio

    2014-12-01

    Full Text Available When data are affected by multicollinearity in the linear regression framework, then concurvity will be present in fitting a generalized additive model (GAM. The term concurvity describes nonlinear dependencies among the predictor variables. As collinearity results in inflated variance of the estimated regression coefficients in the linear regression model, the result of the presence of concurvity leads to instability of the estimated coefficients in GAMs. Even if the backfitting algorithm will always converge to a solution, in case of concurvity the final solution of the backfitting procedure in fitting a GAM is influenced by the starting functions. While exact concurvity is highly unlikely, approximate concurvity, the analogue of multicollinearity, is of practical concern as it can lead to upwardly biased estimates of the parameters and to underestimation of their standard errors, increasing the risk of committing type I error. We compare the existing approaches to detect concurvity, pointing out their advantages and drawbacks, using simulated and real data sets. As a result, this paper will provide a general criterion to detect concurvity in nonlinear and non parametric regression models.

  19. 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

  20. 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.)

  1. 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

  2. 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)

  3. 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

  4. 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)

  5. 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)

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.)

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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)

  4. Environmental information systems - practicable decision aids. Umweltinformationssysteme - praktikable Entscheidungshilfen

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

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

  5. 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

  6. Application of the Vertex Exchange Method to estimate a semi-parametric mixture model for the MIC density of Escherichia coli isolates tested for susceptibility against ampicillin.

    Science.gov (United States)

    Jaspers, Stijn; Verbeke, Geert; Böhning, Dankmar; Aerts, Marc

    2016-01-01

    In the last decades, considerable attention has been paid to the collection of antimicrobial resistance data, with the aim of monitoring non-wild-type isolates. This monitoring is performed based on minimum inhibition concentration (MIC) values, which are collected through dilution experiments. We present a semi-parametric mixture model to estimate the entire MIC density on the continuous scale. The parametric first component is extended with a non-parametric second component and a new back-fitting algorithm, based on the Vertex Exchange Method, is proposed. Our data example shows how to estimate the MIC density for Escherichia coli tested for ampicillin and how to use this estimate for model-based classification. A simulation study was performed, showing the promising behavior of the new method, both in terms of density estimation as well as classification.

  7. Action plan Fukushima 2013; Aktionsplan Fukushima 2013

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-02-15

    As a direct consequence of the accident at the nuclear power plant (NPP) at Fukushima Dai-ichi in Japan in March 2011, the Swiss Federal Nuclear Safety Inspectorate (ENSI) opened measures to check the safety of the Swiss NPPs. In the first 3 orders the operators of the Swiss 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 on the basis of first insights gained from the accident in Japan were also requested. Parallel to the operators' checks ENSI conducted inspections in order to detect eventual weaknesses in the cooling systems of the fuel element storage pools, the protection against external flooding and the systems for filtered depressurization of the containment. The inspections were pursued in 2012 and included the plant strategy in the case of a long-term loss of the power supply, the process and guides for the evaluation of external events and the available emergency premises in the Swiss NPPs. The results of the ENSI's inspections have confirmed that the Swiss NPPs show a high protection against the effects of earthquakes, flooding and combination thereof, and that appropriate measures against the loss of power supply and heat sink have been taken. All analysed accidents can be mastered under taking into account the actual safety measures. So the legal requirements for the respect of the protection goals, i.e. reactivity control, cooling of the fuel elements and enclosing of radioactive materials, are guaranteed. In view to an improvement of the

  8. SEMİPARAMETRİK REGRESYON

    Directory of Open Access Journals (Sweden)

    Münevver TURANLI

    2011-01-01

    Full Text Available SEMIPARAMETRIC REGRESSIONAbstract: Classical (parametric regression techniques are based on the assumption that the independent variable is correlated linearly with the dependent variables and the pattern of this relation is known. When such assumption cannot be verified, parameter estimations fail to be reliable. In cases where the way of correlation is not known or it does not comply with the known parametric mathematical patterns, nonparametric regression techniques are to be applied. One shortcoming concerning this procedure emerges particularly in the interpretation process due to problems brought about by multidimensional aspect of the existence of more than one independent variable. Whenever confronted with a case that includes more than one independent variable, some of the independent variables correlate linearly with the dependent variable; at other times some of the independent variables might correlate nonlinearly. In order to establish a modeling for such relations, semiparametric regression models, comprising the aggregate of parametric and nonparametric regression function, are utilized. İn this study semiparametric regression definitions, estimation (backfitting algorithm, confidence bands, calculating standard errors and hypothesis testing are explained.Keywords: Additive Models, Semiparametric Regression, Backfitting Algorithm.SEMİPARAMETRİK REGRESYONÖzet: Klasik (parametrik regresyon teknikleri, bağımlı değişkenin bağımsız değişkenlerle doğrusal bir ilişki içerisinde olduğunu ve ilişkinin şeklinin biliniyor olduğunu varsayar. Bu varsayımların sağlanamaması durumunda ise parametre tahminleri güvenilir olmamaktadır. İlişkinin şeklinin bilinmediği ya da bilinen parametrik matamatiksel kalıplara uymadığı durumlarda parametrik olmayan regresyon teknikleri kullanılmaktadır. Ancak bu teknikler birden fazla bağımsız değişken olma durumunda çok boyutluluğun yarattığı sıkıntı nedeniyle

  9. 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

  10. 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.)

  11. 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.

  12. 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.

  13. 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.)

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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

  2. 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.)

  3. 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.)

  4. 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.

  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. 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

  8. 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

  9. 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

  10. 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.)

  11. 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.)

  12. 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)

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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)

  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. 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)

  11. 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

  12. Subspace-based additive fuzzy systems for classification and dimension reduction

    Science.gov (United States)

    Jauch, Thomas W.

    1997-10-01

    In classification tasks the appearance of high dimensional feature vectors and small datasets is a common problem. It is well known that these two characteristics usually result in an oversized model with poor generalization power. In this contribution a new way to cope with such tasks is presented which is based on the assumption that in high dimensional problems almost all data points are located in a low dimensional subspace. A way is proposed to design a fuzzy system on a unified framework, and to use it to develop a new model for classification tasks. It is shown that the new model can be understood as an additive fuzzy system with parameter based basis functions. Different parts of the models are only defined in a subspace of the whole feature space. The subspaces are not defined a priori but are subject to an optimization procedure as all other parameters of the model. The new model has the capability to cope with high feature dimensions. The model has similarities to projection pursuit and to the mixture of experts architecture. The model is trained in a supervised manner via conjugate gradients and logistic regression, or backfitting and conjugate gradients to handle classification tasks. An efficient initialization procedure is also presented. In addition a technique based on oblique projections is presented which enlarges the capabilities of the model to use data with missing features. It is possible to use data with missing features in the training and in the classification phase. Based on the design of the model, it is possible to prune certain basis functions with an OLS (orthogonal least squares) based technique in order to reduce the model size. Results are presented on an artificial and an application example.

  13. 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

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. Multiple Regime Semiparametric Smooth Transition Regression Model and Discussion on Business Cycle of China%多机制半参数平滑转换回归模型——兼论我国宏观经济运行周期

    Institute of Scientific and Technical Information of China (English)

    王成勇

    2012-01-01

    本文结合多机制平滑转换回归模型和半参数平滑转换回归模型,提出多机制半参数平滑转换回归模型。对模型转换函数中的未知光滑有界函数采用级数估计,并给出了结合Back—fitting算法和非线性最小二乘法估计模型参数的具体执行步骤,随机模拟结果说明了本文模型和估计算法的可行性和灵活性。应用本文模型和估计算法对我国宏观经济运行周期的实证研究表明,我国经济增长的非线性结构可以分为四个显著不同的增长机制:扩张阶段、衰退阶段、收缩阶段、恢复阶段,并且宏观经济政策的作用有三到四个季度的迟滞效应。%Based on multiple regime smooth transition regression model and semiparametric smooth transition regression model, we proposed the nmltiple regime semiparametrie smooth transition regression model. Series estimator is introduced to estimate the unknown bounded smooth function which contained in smooth transition function, the implementation steps of our estimation algorithm which combined with back-fitting algorithm and NLS are also given. The simulation study indicated that the new model and the estimation algorithm are feasible and flexible. Applied the new model and the estimation algorithm to the business cycle of China, it found that the growing nonlinear structure of China economy can be divided into four distinct regimes: strengthen expansion phase; recession phase; worsening contraction phase; recovery phase, and the adjustment of macroeconomic policy has three or four quarters lag.

  11. 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.)

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

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

    International Nuclear Information System (INIS)

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

  8. 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