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

International Nuclear Information System (INIS)

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

1995-05-01

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

Influence of regulatory requirements for nuclear power plants on the backfitting of Austrian research reactors

International Nuclear Information System (INIS)

Boeck, H.; Hammer, J.

1985-01-01

In general the licensing and backfitting activities have once more demonstrated the fact that safety assessment of a research reactor is by no means just a scaled-down version of a nuclear power plant licensing procedure. Naturally the risk potential is much lower, however, the very nature of research calls for much more flexibility in operation, for temporary installations and for experimental methods which cannot be covered by detailed regulations in advance. Therefore the application of nuclear power reactor criteria to such facilities has to be considered with extreme caution. If NPP standards are applicable at all, they have to be carefully interpreted in each individual case. It is interesting to compare the original reactor safety reports with their modern versions: emphasis has shifted from reactivity accident calculations to thermal-hydraulic considerations, to better instrumentation (both in quality and quantity) and to more effort in reducing, measuring and documenting all radioactive effluents. This tendency is also reflected in most of the backfitting requirements. In summary, the result of the lengthy licensing and backfitting process is certainly a considerable improvement in performance and safety of the Austrian research reactors

Sequence Coding and Search System Backfit Quality Assurance Program Plan

International Nuclear Information System (INIS)

Lovell, C.J.; Stepina, P.L.

1985-03-01

The Sequence Coding and Search System is a computer-based encoding system for events described in Licensee Event Reports. This data system contains LERs from 1981 to present. Backfit of the data system to include LERs prior to 1981 is required. This report documents the Quality Assurance Program Plan that EG and G Idaho, Inc. will follow while encoding 1980 LERs

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

International Nuclear Information System (INIS)

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

1995-05-01

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

NPP backfitting and services

International Nuclear Information System (INIS)

Pauli, W.; Blumer, U.; Edelmann, X.

1988-01-01

The Swiss nuclear power stations are not only of different design (PWR, BWR), but also have been built by several system suppliers over the last 20 years. Consequently, an individual approach has to be adopted for the backfitting and services for each of the five units in operation. Part of the ongoing maintenance effort for nuclear power plants currently in operation is the idea to extend their operating lifetime as long as it appears economic. Although many of the components have been designed for a life of 40 years, their operating period may be extended considerably, provided that proper measures are taken to keep the plants in optimal conditions for this goal. Such measures for a plant life extension program are: assessment of records of actual operating history; additional monitoring where necessary for temperatures, stress, strain, etc.; non-destructive examination techniques; destructive examination of samples removed from service; thermohydraulic and structural analysis for fatigue and fracture; preventive maintenance; repairs; replacement with components of better design and/or material; installation of additions and redundancies where safety is important. 2 refs

Backfitting in Smoothing Spline Anova, with Application to Historical Global Temperature Data

Science.gov (United States)

Luo, Zhen

In the attempt to estimate the temperature history of the earth using the surface observations, various biases can exist. An important source of bias is the incompleteness of sampling over both time and space. There have been a few methods proposed to deal with this problem. Although they can correct some biases resulting from incomplete sampling, they have ignored some other significant biases. In this dissertation, a smoothing spline ANOVA approach which is a multivariate function estimation method is proposed to deal simultaneously with various biases resulting from incomplete sampling. Besides that, an advantage of this method is that we can get various components of the estimated temperature history with a limited amount of information stored. This method can also be used for detecting erroneous observations in the data base. The method is illustrated through an example of modeling winter surface air temperature as a function of year and location. Extension to more complicated models are discussed. The linear system associated with the smoothing spline ANOVA estimates is too large to be solved by full matrix decomposition methods. A computational procedure combining the backfitting (Gauss-Seidel) algorithm and the iterative imputation algorithm is proposed. This procedure takes advantage of the tensor product structure in the data to make the computation feasible in an environment of limited memory. Various related issues are discussed, e.g., the computation of confidence intervals and the techniques to speed up the convergence of the backfitting algorithm such as collapsing and successive over-relaxation.

Seismic hazard assessment in intra-plate areas and backfitting

International Nuclear Information System (INIS)

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

2001-01-01

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

The Alara principle in backfitting Borssele

International Nuclear Information System (INIS)

Leurs, C.J.

1998-01-01

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

Concept for backfitting of earth connections and lightning arresters in accordance with KTA 2206

International Nuclear Information System (INIS)

Kronauer, P.

1991-01-01

Instrumentation and control systems are particularly endangered by overvoltage caused by lightning. Protective aim and scope of the measures to be taken are laid down in the draft regulation KTA 2206 'Design of nuclear power plants against lightning impacts'. In the following a concept is presented which, if implemented, helps to avoid, to a large extent, inadmissible lightning effects on instrumentation and control systems of NPPs, by means of graduated measures of external and internal lightning protection. In the past, this concept was used successfully, in particular with regard to the backfitting of earth connections and lightning arresters of NPPs. (orig./DG) [de

PSA based plant modifications and back-fits

International Nuclear Information System (INIS)

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Brown, R.V.; Ulvila, J.W.

1988-06-01

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

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

International Nuclear Information System (INIS)

Brognon, T.

1993-01-01

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

NPP Evaluation, backfitting and life extension. An engineering viewpoint

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Lopez, A [Empresarios Agrupados, A.I.E., Madrid (Spain)

1993-12-15

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

NPP Evaluation, backfitting and life extension. An engineering viewpoint

International Nuclear Information System (INIS)

Gonzalez Lopez, A.

1993-01-01

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

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

International Nuclear Information System (INIS)

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

1980-01-01

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

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

International Nuclear Information System (INIS)

Fasko, P.

1985-01-01

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

ESTIMACIÓN ROBUSTA DE MODELOS ADITIVOS MEDIANTE EL ALGORITMO DE BACKFITTING

Directory of Open Access Journals (Sweden)

Luis P. Yapu Quispe

2012-07-01

Full Text Available En este trabajo se presenta un método de estimación y simulación de un modelo aditivo a dos variables mediante splines robustos, el método general puede ser aplicado con varias variables. El software utilizado para las simulaciones es S+ y se utiliza explícitamente la función smooth.splineRob en una implementación del algoritmo de backfitting. La función smooth.splineRob ha sido escrita en base al trabajo de Cantoni y Ronchetti [3], en el cual se pone énfasis en la selección robusta del parámetro de suavizamiento utilizando una versión robusta del Cp de Mallows, RCp, y de la validación cruzada, RCV. La existencia de datos extremos o no-normales en la parte estocástica de un modelo aditivo puede provocar una mala estimación del parámetro de suavizamiento, lo que tendrá influencia global en la estimación por splines. Para la etapa de simulación se realizan las estimaciones por splines clásicos y robustos (con estimación robusta del parámetro. La estimación obtenida es muy convincente pero el tiempo de ejecución del programa es relativamente elevado tanto para RCp y RCV, aun cuando, en ciertos casos, con pocas iteraciones robustas se obtienen ya resultados más útiles que la estimación clásica.

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

International Nuclear Information System (INIS)

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

1990-01-01

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

78 FR 40776 - Issuance of Regulatory Guide 1.124 and Regulatory Guide 1.130

Science.gov (United States)

2013-07-08

... constitute backfitting as defined in 10 CFR 50.109 (the Backfit Rule) and is not otherwise inconsistent with... the availability of information regarding these documents. You may access information related to these... section of this document. NRC's Agencywide Documents Access and Management System (ADAMS): You may access...

77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Science.gov (United States)

2012-09-11

... backfitting as defined in 10 CFR 50.109 (the Backfit Rule) and is not otherwise inconsistent with the issue... licenses or combined licenses. This regulatory guide may be applied to applications for operating licenses... may access information related to this document, which the NRC possesses and are publicly available...

78 FR 31614 - Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles

Science.gov (United States)

2013-05-24

... applicants. Issuance of this final ISG may be viewed as constituting backfitting as defined in 10 CFR 50.109 (the Backfit Rule) and may be considered to be inconsistent with the issue finality provisions in 10...: LWR Edition.'' This ISG applies to reviews of ESP, DC and COL applications where the resolution of the...

78 FR 64030 - Monitoring Criteria and Methods To Calculate Occupational Radiation Doses

Science.gov (United States)

2013-10-25

... constitute backfitting as defined in 10 CFR 50.109 (the Backfit Rule) and would not be otherwise inconsistent... licensing basis for the facility. If this draft regulatory guide is finalized, the NRC may apply the revised... improvements in all published guides are encouraged at any time. ADDRESSES: You may submit comment by any of...

The nuclear regulatory challenge of judging safety back fits

International Nuclear Information System (INIS)

2002-01-01

The economic pressures of electricity market competition have led nuclear power plant operators to seek ways to increase electricity production and to reduce operating costs at their plants. Corresponding pressures on the regulatory bodies include operator demand to reduce regulatory burdens perceived as unnecessary and general resistance to consider safety back-fits sought by the regulator. The purpose of this report is to describe potential situations giving rise to safety back-fit questions and to discuss regulatory approaches for judging the back-fits. The intended audience for this report is primarily nuclear regulators, although the information and ideas may also be of interest to nuclear operating organisations, other industry organisations and the general public. (author)

Future control room design (modernization of control room systems)

International Nuclear Information System (INIS)

Reischl, Ludwig; Freitag, Timo; Dergel, Rene

2009-01-01

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

Future control room design (modernization of control room systems); Zukuenftiges Wartendesign (Modernisierung von Warteneinrichtungen)

Energy Technology Data Exchange (ETDEWEB)

Reischl, Ludwig; Freitag, Timo; Dergel, Rene [AREVA NP (Germany). NLLR-G ' ' Reactor I and C' '

2009-07-01

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

Safety analysis of the Morsleben radioactive waste repository (ERAM)

International Nuclear Information System (INIS)

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

1991-03-01

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

VVER-1000 backfitting programs

International Nuclear Information System (INIS)

Zabka, H.; Milhem, J.L.

1998-01-01

Russia, Ukraine, and Bulgaria have nineteen nuclear generating units of the VVER-1000/V-320 (1000 MWe PWR) type in operation. Most of these plants were built in the eighties. Their design is based on Soviet standards of the seventies. In the early eighties and, in particular, after the Chernobyl accident, new safety principles and supplementary specific standards were introduced. However, they were taken into account only to a limited extent in the design and construction of the VVER-1000/V-320 plants. A number of nuclear power plants, whose construction was stopped after the political changes in the countries of the former USSR, now are to be completed with the financial assistance of the Commission of the European Union and other Western organizations, respectively. This Western support is dependent on the condition that these plants attain a level of engineered safeguards comparable to that of PWR plants currently in operation in Western Europe. (orig.) [de

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

International Nuclear Information System (INIS)

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

1979-12-01

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

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

International Nuclear Information System (INIS)

Raetzke, C.

2006-01-01

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

Improvements in the nuclear power plants - a permanent task for the plant management

International Nuclear Information System (INIS)

Langetepe, G.

1991-01-01

The main motives of the operators of nuclear power plants for carrying out backfitting measures are given by the following objectives: (1) to operate the nuclear power plants, older ones too, at a high level of safety, and to keep a lowest possible difference to the respective level of science and technology, (2) to ensure preconditions for a best possible economical operation, also with changing cost structures, (3) to create preconditions for the longest possible operational time. Operational times of more than 40 years have been throught to be realistic. A constant analysis of the operational safety of the whole plant is necessary for laying down the measures for backfitting of the plant. This analysis must also include the valuation of the nuclear safety in accordance with the progressive level of sience and technology. The proccess of backfitting in the German nuclear power plants will be illustrated with the help of several examples. (orig.) [de

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

Energy Technology Data Exchange (ETDEWEB)

Steffen, W. [Federal Office of Energy, Swiss Federal Nuclear Safety Inspectorate, CH-5303 Wuerenlingen (Switzerland)

1986-02-15

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)

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

International Nuclear Information System (INIS)

Steffen, W.

1986-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Shaukat, S K; Jackson, J E; Thatcher, D F

1991-04-01

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

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.

Safety in Swiss nuclear power plants

International Nuclear Information System (INIS)

Cederqvist, H.

1992-01-01

Safety-related facilities and equipment are continuously backfitted in Swiss nuclear power plants. In the Beznau-1 and -2 nuclear generating units, the measures taken under the heading of 'Backfitting of Emergency Systems' included provisions to enhance the protection against earthquakes, airplane crash, and fire; in addition, the emergency power system was upgraded. In Muehleberg, the stack exhaust air monitoring system was optimized. The containment pressure suppression system of the plant has been designed to withstand a hypothetical accident exceeding the design basis. The BKM-Crud computer simulation model simulates steps taken to reduce radiation exposure. The power of Swiss nuclear power stations will be raised by 4% to 15% within the 'Energy 2000' action program. (orig.) [de

Plant upgrading and backfitting in France

International Nuclear Information System (INIS)

Moxley, Nigel; Raimondo, Emile

1992-01-01

The service life of a nuclear power plant is linked to good operating and maintenance practices and effective management which should begin when the plant starts operating. In France the construction of two series of virtually identical reactors has enabled Framatome to use feedback from early units to modify and improve later ones. The intention is to keep the plants safe and to extend life as far into the next century as is feasible. Remedial measures that can be applied include improvements to plant operation itself, plant modifications and the implementation of effective preventive and corrective maintenance operations which may cover planned provision for the replacement of obsolete or susceptible components. (author)

Operating experience with Beloyarsk fast reactor BN600 NPP

International Nuclear Information System (INIS)

Saraev, O.M.

2000-01-01

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

Nonparametric additive regression for repeatedly measured data

KAUST Repository

Carroll, R. J.; Maity, A.; Mammen, E.; Yu, K.

2009-01-01

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

Passive systems for light water reactors

International Nuclear Information System (INIS)

Adinolfi, R.; Noviello, L.

1990-01-01

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

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

International Nuclear Information System (INIS)

1993-01-01

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

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

International Nuclear Information System (INIS)

Versteeg, M.F.; Vos, D.

1994-01-01

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

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

International Nuclear Information System (INIS)

1993-01-01

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

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.

78 FR 7816 - Quality Assurance Program Requirements (Operations)

Science.gov (United States)

2013-02-04

... defined in 10 CFR 50.109 (the Backfit Rule) and is not otherwise inconsistent with the issue finality... or combined licenses. This regulatory guide may be applied to applications for operating licenses and... may access information and comment submissions related to this document, which the NRC possesses and...

78 FR 48503 - Proposed Revision to Missiles Generated by Extreme Winds

Science.gov (United States)

2013-08-08

... constitute backfitting as defined in 10 CFR 50.109, or otherwise be inconsistent with the issue finality.... ADDRESSES: You may submit comments by any of the following methods (unless this document describes a.... You may access information related to this document, which the NRC possesses and is publicly available...

77 FR 59023 - Preoperational Testing of Instrument and Control Air Systems

Science.gov (United States)

2012-09-25

... regulatory guide may be applied to applications for operating licenses and combined licenses docketed by the... does not constitute backfitting as defined in 10 CRF 50.109(a)(1) or is otherwise inconsistent with the... of information regarding this document. You may access information related to this document, which...

78 FR 41810 - Proposed Revisions to Light Load Handling System and Operations

Science.gov (United States)

2013-07-11

... constitute backfitting as defined in 10 CFR 50.109, or otherwise be inconsistent with the issue finality... on or before this date. ADDRESSES: You may submit comments by any of the following methods (unless... of information regarding this document. You may access information related to this document, which...

78 FR 41434 - Proposed Revisions to Design of Structures, Components, Equipment and Systems

Science.gov (United States)

2013-07-10

... backfitting as defined in 10 CFR 50.109, or otherwise be inconsistent with the issue finality provisions in 10.... ADDRESSES: You may submit comments by any of the following methods (unless this document describes a.... You may access information related to this document, which the NRC possesses and is publicly available...

Smolensk gets GOMIS

International Nuclear Information System (INIS)

Dynan, John; Francis, Arthur

1993-01-01

Improving safety at the Soviet designed RBMK reactors is not simply a matter of backfits. Effective management information systems and maintenance schedules will be needed, and not just for increased safety. A computer-based maintenance management information system is described, called GOMIS as are the introduction of various Nondestructive Testing programs. (Author)

Institutionalization of safety re-assessment system for operating nuclear power plants

International Nuclear Information System (INIS)

Kim, H. J.; Cho, J. C.; Min, B. K.; Park, J. S.; Jung, H. D.; Oh, K. M.; Kim, W. K.; Lim, J. H.

1999-01-01

In this study, in-depth reviews of the foreign countries' experiences and practices in applications of the periodic safety review (PSR), backfitting and license renewal systems as well as the current status of nuclear power safety assurance programs and activities in Korea have been performed to investigate the necessity and feasibility of the application of the systems for the domestic operating nuclear power plants and to establish effective strategy and methodology for the institutionalization of a periodic safety re-assessment system appropriate to both the domestic and international nuclear power environments by incorporating the PSR with the backfitting and license renewal systems. For these purposes, the regulatory policy, fundamental principles and detailed requirements for the institutionalization of the safety re-assessment system and the effective measures for active implementation of the backfitting program have been developed and then a comparative study of benefits and shortcomings has been conducted for the three different models of the periodic safety re-assessment system incorporated with either the license renewal or life extension process, which have been considered as practicable ones in the domestic situation. The model chosen in this study as the most appropriate safety re-assessment system is the one that the re-assessments are performed at the interval of ten years throughout the service life of nuclear power plant and the ten-year license renewal or life extension after the expiration of design life can be permitted based on the regulatory review of the re-assessment results and follow-up measures. Finally, this paper has discussed on the details of the requirements, approach and procedures established for the institutionalization of the periodic safety re-assessment system chosen as the most appropriate one for domestic applications

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.

AMNT 2014. Key Topic: Reactor operation, safety - report. Pt. 1

International Nuclear Information System (INIS)

Schaffrath, Andreas

2014-01-01

Summary report on one session of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Safety of Nuclear Installations - Methods, Analysis, Results: Backfittings for the Improvement of Safety and Efficiency. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' will be covered in further issues of atw.

LWR design decision methodology: Phase II. Final report

International Nuclear Information System (INIS)

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

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

International Nuclear Information System (INIS)

Anon.

1994-01-01

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

Safety upgrade at the Leningrad NPP

International Nuclear Information System (INIS)

Eperin, A.P.

1996-01-01

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

Improved sealing for in-core systems

International Nuclear Information System (INIS)

Dunford, S.

1989-01-01

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

Emission and thermal performance upgrade through advanced control backfit

Energy Technology Data Exchange (ETDEWEB)

Banerjee, A.K. [Stone & Webster Engineering Corporation, Boston, MA (United States)

1994-12-31

Reducing emission and improving thermal performance of currently operating power plants is a high priority. A majority of these power plants are over 20 years old with old control systems. Upgrading the existing control systems with the latest technology has many benefits, the most cost beneficial are the reduction of emission and improving thermal performance. The payback period is usually less than two years. Virginia Power is installing Stone & Webster`s NO{sub x} Emissions Advisor and Advanced Steam Temperature Control systems on Possum Point Units 3 and 4 to achieve near term NO{sub x} reductions while maintaining high thermal performance. Testing has demonstrated NO{sub x} reductions of greater than 20 percent through the application of NO{sub x} Emissions Advisor on these units. The Advanced Steam Temperature Control system which has been operational at Virginia Power`s Mt. Storm Unit 1 has demonstrated a signification improvement in unit thermal performance and controllability. These control systems are being combined at Units 3 and 4 to reduce NO{sub x} emissions and achieve improved unit thermal performance and control response with the existing combustion hardware. Installation has been initiated and is expected to be completed by the spring of 1995. Possum Point Power Station Units 3 and 4 are pulverized coal, tangentially fired boilers producing 107 and 232 MW and have a distributed control system and a PC based performance monitoring system. The installation of the advanced control and automation system will utilize existing control equipment requiring the addition of several PCs and PLC.

Balancing safety and economics

International Nuclear Information System (INIS)

Kroeger, W.; Fischer, P.U.

2000-01-01

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

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

International Nuclear Information System (INIS)

Ishii, Hirofumi

1993-01-01

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

Upgrading the safety of VVER-440/V-230

International Nuclear Information System (INIS)

Kelm, P.; Wenk, W.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1988-07-01

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

AMNT 2014. Key Topic: Reactor operation, safety - report. Pt. 1

Energy Technology Data Exchange (ETDEWEB)

Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany). Forschungszentrum

2014-10-15

Summary report on one session of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Safety of Nuclear Installations - Methods, Analysis, Results: Backfittings for the Improvement of Safety and Efficiency. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' will be covered in further issues of atw.

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)

Mixture of Regression Models with Single-Index

OpenAIRE

Xiang, Sijia; Yao, Weixin

2016-01-01

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

Investigation of WWER fuel behaviour under MIR power ramps

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Novikov, V.V.; Agafonov, S.N.

1996-01-01

The paper discusses results of experimental WWER fuel investigation under power ramps. Specificity of using the research reactor ''MIR'' to accomplish scheduled power rating of fuel is considered. The paper presents the methodology of experiments using irradiation facility ''TEST''. Reactor experiments were performed at burn-up ∼ 10000 MW.day/t UO 2 using standard fuel pins and the ones having backfitted fuel and cladding. (author). 7 figs, 1 tab

Distributed software applications in JAVA for portable processors operating on a wireless LAN

OpenAIRE

Rothenhaus, Kurt J.

1999-01-01

Approved for pubic release; distribution is unlimited As the wave of future Information Technology makes its way into the construction and design of new ships and submarines, it is imperative to examine methods to thoroughly economically backfit older platforms with similar technology. Affordable, Commercial-off-the-shelf (COTS) industrial products have provided us with a means to reduce miscommunication and exponentially increase the availability of information via small pen based compute...

Regulatory licensing, status summary report; Systematic evaluation program

International Nuclear Information System (INIS)

1978-01-01

This document is part of a management information system presenting a logical flow of events that represent the evaluation of 11 of the older operating nuclear reactors. Information collected will be used to determine the degree to which the 11 plants meet current licensing requirements and to develop an overall balanced position concerning any needed backfitting of the facilities and the documentation of the results of such evaluations

Modernizing the VVER-440/230

International Nuclear Information System (INIS)

Mink, F.J.

1991-01-01

The modernization of the VVER-440/230s is not fundamentally different from backfit projects on older pressurized water reactors which Westinghouse has completed elsewhere. However, carrying out such programmes only makes sense if the plants are expected to continue operation for their projected life or beyond. This clearly requires some licensing and political stability; both are essential if investors in the upgrading project are to be found. (author)

Control rooms in German nuclear power plants

International Nuclear Information System (INIS)

Hoffmann, E.

1999-01-01

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

Modernisation of the Borssele NPP reactor protection system

International Nuclear Information System (INIS)

Plas, Y. van der

1993-01-01

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

Ergonomics: an aid to system design

International Nuclear Information System (INIS)

McCafferty, D.B.

1990-01-01

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

Oskarshamn 1-project FENIX

International Nuclear Information System (INIS)

Sjoeqvist, N.G.

1994-01-01

This paper summarizes the actions to be taken in a large re-start and backfitting project such as the Fenix project. It describes the organization, planning and financial management, safety criteria and licensing procedures, safety concept report, health and safety. The results from the unique full system decontamination of the reactor pressure vessel is described. The project is still ongoing and therefore other results and lessons learnt are not reported. (author) 9 figs

Backfitting of Nuclear Power Plant Bohunice V1 in Slovakia

International Nuclear Information System (INIS)

Ferenc, M.

1999-01-01

Nuclear power plants in the Slovak Republic generate almost 55 % of electricity. The operating organization and the Nuclear Regulatory Authority of the Slovak Republic pay a great attention to safe and reliable operation of four units with VVER 440 reactors at Bohunices site and one in Mochovce side. Engineering and design organizations in cooperation with well known international companies prepare evaluation of safety conditions, safety analyses and projects for the implementation of modifications to upgrade the nuclear safety of the units in operation. A gradual safety upgrading (reconstruction) of the V-1 Bohunice plant has been in progress, a modernization of the V-2 Bohunice plant is being prepared. Simultaneously the commissioning of Unit 2 at the Mochovce plant is being implemented.(author)

Backfitting of the nuclear plant V1 power control system

International Nuclear Information System (INIS)

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

1985-01-01

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

FRAMATOME nuclear services

International Nuclear Information System (INIS)

Delorme, H.; Buttin, J.

1985-05-01

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

Belgian national report

International Nuclear Information System (INIS)

Berthe, J.

1995-01-01

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

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

International Nuclear Information System (INIS)

Brooksbank, R.E. Sr.

1980-01-01

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

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

International Nuclear Information System (INIS)

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

1988-07-01

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

Combustion Engineering adjusts to slump in nuclear orders

International Nuclear Information System (INIS)

Masters, R.

1980-01-01

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

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

International Nuclear Information System (INIS)

Lipar, M.; Kerak, J.; Rohar, S.

1998-01-01

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

Ignalina Nuclear Power Plant - the management of a major backfitting exercise

International Nuclear Information System (INIS)

Bissell, J.

1995-01-01

This paper describes the work being undertaken at Ignalia NPP on the Nuclear Safety Project using funds managed by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). The background of the Nuclear Safety Account and the Grants made from it are summarised. Finally the work of the Project Management Unit (PMU) is described and lessons learned are outlined. (author)

Operating cost reduction by optimization of I and C backfitting strategy

International Nuclear Information System (INIS)

Kraft, Heinz-U.

2002-01-01

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

Guidelines for control room systems design. Working material. Report

International Nuclear Information System (INIS)

1993-01-01

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

Coal handling system structural analysis for modifications or plant life extension

International Nuclear Information System (INIS)

Dufault, A.; Weider, F.; Doyle, P.

1989-01-01

One neglected aspect of plant modification or life extension is the extent to which previous projects may have affected the integrity of existing structures. During the course of a project to backfit fire protection facilities to existing coal handling systems, it was found that past modifications had added loads to existing coal handling structures which exceeded the available design margin. This paper describes the studies that discovered the original problem areas, as well as the detailed analysis and design considerations used to repair these structures

Steam generator replacement at Doel 3 NPP (Belgium)

International Nuclear Information System (INIS)

Danhier, B.

1993-01-01

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

Partner of nuclear power plants

International Nuclear Information System (INIS)

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

1992-01-01

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

Risk - Informed decision making at Loviisa NPP

International Nuclear Information System (INIS)

Vaurio, J.K.

1999-01-01

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

Use of digital photography for power plant retrofits

International Nuclear Information System (INIS)

Kamba, J.J.

1995-01-01

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

Risk - Informed decision making at Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

Vaurio, J.K. [Fortum Power and Heat Oy, Loviisa (Finland)

1999-09-01

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

AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 2

International Nuclear Information System (INIS)

Fischer, Klaus-Christian; Willschuetz, Hans-Georg; Wortmann, Birgit

2014-01-01

Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Thermo Dynamics and Fluid Dynamics: Experiments and Backfittings for the Improvement of Safety and Efficiency; - Safety of Nuclear Installations - Methods, Analyses, Results: In-Vessel Phenomena; Ex-Vessel Phenomena; - Standards and Regulations; Hazard and Safety Analysis; and Validation and Uncertainty Analysis. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 (2014) and will be covered in further issues of atw.

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

International Nuclear Information System (INIS)

Feuerbach, R.

2001-01-01

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

Acoustically damped metal oil trough for internal combustion engines. Schallgedaempfte Blech-Oelwanne fuer Brennkraftmaschinen

Energy Technology Data Exchange (ETDEWEB)

Kubis, H.

1991-03-28

The invention refers to an acoustically damped oil trough. As there are strict requirements for reducing the noise emission from internal combustion engines, according to the invention it is proposed that the oil trough should be surrounded by an outer trough, where the outer trough is made of plastic or sheet steel in one or more layers. To avoid noise bridges, the oil trough and outer trough are separated by elastomer elements. The outer trough achieves a reasonably priced increase in sound insulation. It is also possible to backfit an outer trough on engines.

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

International Nuclear Information System (INIS)

Monette, P.

1995-01-01

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

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

International Nuclear Information System (INIS)

1990-06-01

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

Replacement strategy for obsolete plant computers

International Nuclear Information System (INIS)

Schaefer, J.P.

1985-01-01

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

Control room systems design for nuclear power plants

International Nuclear Information System (INIS)

1995-07-01

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

Control room systems design for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

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

Digital I and C for nuclear power plant

International Nuclear Information System (INIS)

Gemst, P. van

1993-01-01

A summary is given of the past experience (process I and C, digital controllers, Power Range Monitoring system) and future plans (integrated plant I and C, control room) of ABB Atom for programmable I and C at nuclear power plants. ABB Atom has designed and supplied an appreciable quantity of software based equipment for nuclear power plants. These have been supplied for both new plants as well as for backfitting. The well proven ABB Master system has been used for the supply of I and C equipment for these projects and will continue to be used in the future. (Z.S.) 1 fig

Working group 5: Safety

International Nuclear Information System (INIS)

Vinck, W.

1976-01-01

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

Management of waste from nuclear facilities as a regulatory problem. Requirements to be met by legislation under conditions of uncertainty. Die Entsorgung der Kernenergie als Regelungsproblem. Zu den Anforderungen an Gesetzgebung unter Ungewissheitsbedingungen

Energy Technology Data Exchange (ETDEWEB)

Ladeur, K.H.

1989-07-01

The author presents a brief review of the development of the nuclear waste management regime in the Atomic Energy Act, referring also to court decissions and the literature. The article analyses the constitutionality of the waste management regulations of section 9a and following sections, and of the provisions on reprocessing (section 7, sub-sec. (1)), primarily under the aspect of the principle of proviso of legality in general, reformulated by the theory of materiality, and in particular with regard to the requirement of 'backfitting' in order to improve the regulatory system for complex and especially technological matters. (orig./RST).

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

International Nuclear Information System (INIS)

Dueymes, E.; Peyrelongue, J.P.

1992-01-01

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

SIEMENS

International Nuclear Information System (INIS)

2001-01-01

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

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

International Nuclear Information System (INIS)

Guicherd, L.

1984-01-01

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

Nuclear power plant control and instrumentation in Switzerland

International Nuclear Information System (INIS)

Voumard, A.

1992-01-01

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

Report on nuclear power plant instrumentation and control in Germany

International Nuclear Information System (INIS)

Bastl, W.

1992-01-01

The paper describes the status of the NPP control and instrumentation in Germany. The general technology underlying most aspects of NPP C and I in Germany has not altered since the last progress report although there has been many improvements in detail. Since the beginning of 1990 the GRS carried out the safety investigations of NPPs in East Germany. The USSR as the vendor of the plants and France were also involved in the project. The following fields are briefly described: Status of nuclear power in Germany; training simulators; backfitting of computers and information systems; operator support/new control rooms. (author). 6 refs, 1 tab

Assessment of policy issues in nuclear safety regulation according to circumstantial changes

Energy Technology Data Exchange (ETDEWEB)

Chang, Soon Heung; Lee, Byong Ho; Baek, Woon Pil; Lee, Seong Wook; Choi, Seong Soo; Roh, Chang Hyun; Lee, Kwang Gu [Korea Advanced Institute of Scienc and Technology, Taejon (Korea, Republic of)

1998-03-15

The objective of the work is to assess various issues in nuclear safety regulation in consideration of circumstantial changes. Emphasis is given to the safety of operating NPPs. The derivation of an effective regulation system considering 'Rhodic Safety Review (PSR)', 'operating License Renewal (LR)', 'backfitting' and 'maintenance rule' is the main objective of the first two years. It is found that those approaches should be introduced in Korea as soon as possible, with cross lingkage to maximize the effectiveness of regulation. In particular, the approaches for PSR are discussed with consultation of IAEA document and foreign practices.

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

International Nuclear Information System (INIS)

1996-01-01

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

Advances in Canadian regulatory practice

International Nuclear Information System (INIS)

Waddington, J.G.

1993-03-01

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

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

International Nuclear Information System (INIS)

Krull, W.

1989-01-01

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

Backfitting of existing nuclear power plants with particulate, iodine and noble gas monitors

International Nuclear Information System (INIS)

Marley, M.R.; Geiger, E.L.

1978-01-01

A stand-alone microcomputer complete with hardware and software to measure airborne particulate iodine and noble gases is described. This system meets the need at power plants and effluent monitoring. The equipment will accommodate up to 192 channels of input

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

International Nuclear Information System (INIS)

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

1993-01-01

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

The V-1 NPP and V-2 NPP upgrading

International Nuclear Information System (INIS)

1998-01-01

A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

SAFARI-1: 30 years of operation

International Nuclear Information System (INIS)

D'Arcy, A.J.; Niebuhr, H.W.; Procter, G.I.

1995-01-01

SAFARI-1, a 20 MW tank-in-pool type MTR, was commissioned in 1965. It today enjoys some benefits from having operated for 17 years at a quarter of its design power. Ageing technology, non-availability of spares and wear and tear are, however, beginning to take their toll, while changing licensing requirements and shifting focus in its utilisation are at the same time compelling the AEC to invest capital in renovation, refurbishment, backfitting and upgrading in several different areas. A brief look at the operating history of SAFARI-1 is followed by a discussion of some of the more important upgrades recently carried out, focussing particular attention on a redesign of the Fail-Free Power System. (orig.)

Radiation protection of workers in mines

International Nuclear Information System (INIS)

1986-01-01

An ICRP report (publication 47) is presented which describes the principles and applications of methods by which radiation hazards may be controlled in mines, particularly in the uranium mining industry. Details are given of the dose limits for individual exposures from 222 Rn, 220 Rn and their decay products and ore dust. The measures described for controlling exposure are choice of mining method, source isolation, mechanical ventilation, air cleaning, backfitting, personal protective equipment and organization of work. Recommendations for air monitoring for radon and radon decay products and ore dust, external exposure monitoring and monitoring the quality of protective measures systems are also presented. Finally, recommendations on medical surveillance of miners are given. (UK)

Licensing, supervision, retrofitting

International Nuclear Information System (INIS)

Steinkemper, H.

1991-01-01

The following proposals for the amendment to the Atomic Energy Act are made: the term of provisions against damage and the content and scope of the principle of commensurability should be defined by law. Their concretization should be left to the level of the statutory instruments and technical codes. In usage the scope of application of the subsequent obligation should be approximated to the category of element relevant to licensing. Lability to indemnification for subsequent obligations should be abolished. The need for a backfitting licence in the case of 'substantial' alterations requires a closer definition. A legal obligation should be placed on operators of nuclear reactors to carry out periodical safety checks. (orig./HSCH) [de

Backfitting of the nuclear power plant Borssele. A practical application of the ALARA-principle

International Nuclear Information System (INIS)

Leurs, C.J.

1998-01-01

In two articles an overview is given of the adjustments for the title plant and how the radiation load of employees is kept as low as possible according to the newest insights of the ALARA-principle. In this article attention is paid to the results of the project 'Modifications' which comprises 16 system engineering modifications. The maximum attention for radiation protection resulted in a collective dose of circa 2500 millisievert, which is a factor 4 lower than the first estimation of the radiation load. 7 refs

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

International Nuclear Information System (INIS)

Leung, V.; Basdekas, D.; Mazetis, G.

1991-06-01

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

Reactor instrumentation and control in nuclear power plants in Germany

International Nuclear Information System (INIS)

Aleite, W.

1993-01-01

The pertinent legislation, guidelines and standards of importance for nuclear power plant construction as well as the relevant committees in Germany are covered. The impact of international developments on the German regulatory scene is mentioned. A series of 15 data sheets on reactor control, followed by 5 data sheets on instrumentation and control in nuclear power plants, which were drawn up for German plants, are compared and commented in some detail. Digitalization of instrumentation and control systems continues apace. To illustrate the results that can be achieved with a digitalized information system, a picture series that documents a plant test of behavior on simulated steam generator tube rupture is elaborately commented. An outlook on backfitting and upgrading applications concludes this paper. (orig.) [de

Nuclear criticality safety department training implementation

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

1991-12-01

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

LOFT Augmented Operator Capability Program

International Nuclear Information System (INIS)

Hollenbeck, D.A.; Krantz, E.A.; Hunt, G.L.; Meyer, O.R.

1980-01-01

The outline of the LOFT Augmented Operator Capability Program is presented. This program utilizes the LOFT (Loss-of-Fluid Test) reactor facility which is located at the Idaho National Engineering Laboratory and the LOFT operational transient experiment series as a test bed for methods of enhancing the reactor operator's capability for safer operation. The design of an Operational Diagnotics and Display System is presented which was backfit to the existing data acquisition computers. Basic color-graphic displays of the process schematic and trend type are presented. In addition, displays were developed and are presented which represent safety state vector information. A task analysis method was applied to LOFT reactor operating procedures to test its usefulness in defining the operator's information needs and workload

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Measures taken to improve nuclear safety on EdF PWRs in operation

International Nuclear Information System (INIS)

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

1993-01-01

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

Use of CEDB for PSA

International Nuclear Information System (INIS)

Balesteri, S.; Besi, A.; Carlesso, S.; Colombo, A.G.; Jaarsma, R.J.

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

International Nuclear Information System (INIS)

1978-05-01

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

Session III: International practice for the requalification of existing NPPs

International Nuclear Information System (INIS)

Haas, E.

1993-01-01

Like in other countries, in Germany, too, a periodic safety reevaluation (Periodische Sicherheitsueberpruefung PSUE) has been recommended to become obligatory for all NPPs. This caused an acceleration of the process of seismic risk evaluation and requalification as a partial aspect of PSUE. For some of the older German NPPs, started up about 1970 to 1980, detailed seismic reevaluations followed by backfitting proposals were performed. In general the seismic acceptance criteria are conform with the demands of the actually valid German standards and regulations. For the methods of evaluation and requalification these standards allow to apply sufficiently variable tools; thus the procedures consist of an adequate combination of analysis, walkdown results, test experience as well as analogue and plausibility considerations. Examples are presented for the main categories of mechanical and electrical equipment of German NPPs

Is there a parliamentary reservation with regard to reprocessing technology?

International Nuclear Information System (INIS)

Rossnagel, A.

1987-01-01

The decision whether the F.R.G. shall commit itself to closing the nuclear fuel cycle and start using plutonium as an energy source may not be taken implicitly, by the licensing competence of the executive organs, but is a problem falling into the area reserved to the decision of Parliament on the basis of a systematic and comprehensive cost-benefit analysis. Because there are novel and essential developments and research results regarding the reprocessing of nuclear fuels, Parliament has to react by 'backfitting' the Atomic Energy Act. Until Parliament has fulfilled this duty, the licensing authorities do not have the competence to grant permits for large-scale processing or storage of plutonium, nor does the Federal Minister of the Interior have any right to instruct the licensing authorities accordingly. (orig./HSCH) [de

Activities on PIRA Researches and its Applications in Korea

International Nuclear Information System (INIS)

Yoo, Kunjoong; Chae, Sungki; Choi, Youngsang; Roh, Myungsub

1986-01-01

Since 1979, activities on the developments of PIRA methodologies and procedures, and its application as well as the safety goal establishment have increased because of the important benefits which may be derived from the uses of PIRA. PIRA is used in many areas including safety evaluation, preparation of basis of standards, siting of nuclear facilities, operator training, energy technology comparison, design, safety improvement(backfitting), operating procedure changes, test and maintenance procedure development, safety goal establishment, and plant availability improvement. Especially, vendor countries are utilizing PIRA methods for parts of the safety demonstration for proposed power plants. In this connection, Korea began to take a interest in PIRA in 1979. Since then, basic methodologies and computer codes of PIRA have been introduced, modified, and applied to Korea nuclear power plants

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Environmental information systems - practicable decision aids

International Nuclear Information System (INIS)

1988-01-01

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

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).

PWR standardization: The French experience

International Nuclear Information System (INIS)

Bacher, P.E.

1987-01-01

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

Nuclear terrorism - an unavoidable companion of nuclear fission?

International Nuclear Information System (INIS)

Mueller, H.

1989-01-01

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) [de

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Nuclear Criticality Safety Department Qualification Program

International Nuclear Information System (INIS)

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

1996-01-01

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

Procedures for maintenance and repairs

International Nuclear Information System (INIS)

Pickel, E.

1981-01-01

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

Probabilistic safety analysis of the Kozloduy NPP units 1-4 (WWER-440/230) using independent emergency feedwater system; Veroyatnostnyj analiz bezopasnosti I-IV blokov AEhS `Kozloduy` s reaktorami tipa WWER-440 (V 230) pri vklyuchenii nezavisimoj sistemy avarijnoj podpitki PG

Energy Technology Data Exchange (ETDEWEB)

Kalchev, B; Marinov, M; Dimitrov, B; Avdzhiev, K [Energoproekt, Sofia (Bulgaria)

1996-12-31

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

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.

The economic impact of reactor transients

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

Kurten, H.

1990-01-01

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

State-of-the-art incore detector system provides operational and safety benefits: Example, Hanford N Reactor

International Nuclear Information System (INIS)

Toffer, H.

1988-08-01

A presentation on the operational and safety benefits that can be derived from a state-of-the-art incore neutron monitoring system has been prepared for the DOE/ANL training course on ''The Potential Safety Impact of New and Emerging Technologies on the Operation of DOE Nuclear Facilities.'' Advanced incore neutron flux monitoring systems have been installed in some commercial reactors and should be considered for any new reactor designs or as backfits to existing plants. The recent installation of such a system at the Hanford N Reactor is used as an example in this presentation. Unfortunately, N Reactor has been placed in a cold standby condition and the full core incore system has not been tested under power conditions. Nevertheless, the evaluations that preceded the installation of the full core system provide interesting insight into the operational and safety benefits that could be expected

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-08-01

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

Probabilistic safety assessment for research reactors

International Nuclear Information System (INIS)

1986-12-01

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

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

International Nuclear Information System (INIS)

1998-08-01

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

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

International Nuclear Information System (INIS)

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

1987-02-01

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

Tsunami hazard

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Seismic and tsunami safety margin assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Analysis of Millstone Unit 1 system failure and maintenance data

International Nuclear Information System (INIS)

Bickel, J.H.; Beveridge, R.L.; Jain, N.K.; Owens, D.B.; Radder, J.A.

1985-01-01

As a result of a task force plan developed four years ago at Northeast Utilities, plant-specific probabilistic safety analysis models are being developed for all Northeast Utilities operating nuclear plants. An essential feature of these models is their reliance on plant-specific reliability information to the maximum extent possible. This assures that future design efforts and decisions on backfitting or procedure changes are made with full knowledge of existing plant reliability. The use of plant-specific reliability data assures that the impacts of problem components are given appropriate attention and that proper credit is given for those components, which because of plant-specific maintenance practices, have exhibited better than industry average performance. A case study of a portion of the Millstone-1 cooling system demonstrates differing results obtained by fault tree analysis and a reliability analysis using plant-specific failure data. When risk assessment techniques are being applied in resource allocation, usage of plant data clearly becomes essential for sound decision making

Peer evaluation and some valuable lessons

Energy Technology Data Exchange (ETDEWEB)

Holt, A G [Ontario Hydro, Toronto, ON (Canada)

1991-04-01

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

Peer evaluation and some valuable lessons

International Nuclear Information System (INIS)

Holt, A.G.

1991-01-01

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

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.

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

International Nuclear Information System (INIS)

Boettcher, D.

1983-01-01

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

Duty and role of Nuclear Regulation Authority facing a crucial moment

International Nuclear Information System (INIS)

Asaoka, Mie

2013-01-01

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

Seismic and tsunami safety margin assessment

International Nuclear Information System (INIS)

2013-01-01

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

Establishment of nuclear knowledge and information infrastructure; establishment of web-based database system for nuclear events

Energy Technology Data Exchange (ETDEWEB)

Park, W. J.; Kim, K. J. [Korea Atomic Energy Research Institute , Taejeon (Korea); Lee, S. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

2001-05-01

Nuclear events data reported by nuclear power plants are useful to prevent nuclear accidents at the power plant by examine the cause of initiating events and removal of weak points in the aspects of operational safety, and to improve nuclear safety in design and operation stages by backfitting operational experiences and practices 'Nuclear Event Evaluation Database : NEED' system distributed by CD-ROM media are upgraded to the NEED-Web (Web-based Nuclear Event Evaluation Database) version to manage event data using database system on network basis and the event data and the statistics are provided to the authorized users in the Nuclear Portal Site and publics through Internet Web services. The efforts to establish the NEED-Web system will improve the integrity of events data occurred in Korean nuclear power plant and the usability of data services, and enhance the confidence building and the transparency to the public in nuclear safety. 11 refs., 27 figs. (Author)

Advances in probabilistic risk analysis

International Nuclear Information System (INIS)

Hardung von Hardung, H.

1982-01-01

Probabilistic risk analysis can now look back upon almost a quarter century of intensive development. The early studies, whose methods and results are still referred to occasionally, however, only permitted rough estimates to be made of the probabilities of recognizable accident scenarios, failing to provide a method which could have served as a reference base in calculating the overall risk associated with nuclear power plants. The first truly solid attempt was the Rasmussen Study and, partly based on it, the German Risk Study. In those studies, probabilistic risk analysis has been given a much more precise basis. However, new methodologies have been developed in the meantime, which allow much more informative risk studies to be carried out. They have been found to be valuable tools for management decisions with respect to backfitting, reinforcement and risk limitation. Today they are mainly applied by specialized private consultants and have already found widespread application especially in the USA. (orig.) [de

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

International Nuclear Information System (INIS)

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

1981-01-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Passive safety features in current and future water cooled reactors

International Nuclear Information System (INIS)

1990-11-01

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

Regulatory analysis technical evaluation handbook. Final report

International Nuclear Information System (INIS)

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Ball, J.R.

1986-04-01

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

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

International Nuclear Information System (INIS)

Tanabe, Tomoyuki; Maruyama, Masahiro

2016-01-01

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

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

International Nuclear Information System (INIS)

Matzie, R.A.

1980-04-01

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

CANDU safety management in Pakistan. A status report

Energy Technology Data Exchange (ETDEWEB)

Mazhar Hasan, S; Badshah Hussain, S; Mirza, K F; Siddiqui, Z H [Karachi Nuclear Power Plant (KANUPP) (Pakistan)

1997-12-01

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

Workshop on environmental qualification of electric equipment

International Nuclear Information System (INIS)

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

1994-05-01

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

CANDU safety management in Pakistan. A status report

International Nuclear Information System (INIS)

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

1997-01-01

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

Improved core monitoring for improved plant operations

International Nuclear Information System (INIS)

Mueller, N.P.

1987-01-01

Westinghouse has recently installed a core on-line surveillance, monitoring and operations systems (COSMOS), which uses only currently available core and plant data to accurately reconstruct the core average axial and radial power distributions. This information is provided to the operator in an immediately usable, human-engineered format and is accumulated for use in application programs that provide improved core performance predictive tools and a data base for improved fuel management. Dynamic on-line real-time axial and radial core monitoring supports a variety of plant operations to provide a favorable cost/benefit ratio for such a system. Benefits include: (1) relaxation or elimination of certain technical specifications to reduce surveillance and reporting requirements and allow higher availability factors, (2) improved information displays, predictive tools, and control strategies to support more efficient core control and reduce effluent production, and (3) expanded burnup data base for improved fuel management. Such systems can be backfit into operating plants without changing the existing instrumentation and control system and can frequently be implemented on existing plant computer capacity

Plant life extension and ageing mechanisms: an ANSALDO proposal for the application to Kozloduy NPPs

International Nuclear Information System (INIS)

Orlandi, S.; Macco, A.; Zanaboni, P.

1999-01-01

In the frame of extension of NPP's installations lifetime, ageing management has become a topical subject to obtain the items: - Evaluation of residual life of the plant through the investigation of the residual life of well identified Safety Related Equipment in the as built configuration; - Organization of identified representative equipment per typological Classes (as Piping Systems, Tanks, Valves, Pumps, Electrical Equipment, Pressurized Components) in order to define for each High Level Class a set of elementary families capable to have, within each family, a common ageing mechanism and methodological investigation and potential common on-line monitoring; Application of a consistent methodology for residual life evaluation at each High Level Class (and subsequent elementary family and group if any) in order to assess an integral approach to backfitting and ageing management, taking also into account the economical investment effort required to the utility. In this report, a technical proposal for the application of the ANSALDO standard methodology approach, applied to the re-evaluation of the Kozloduy NPP's (Unit 1 - to 4) is presented. (authors)

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.

Tsunami hazard

International Nuclear Information System (INIS)

2013-01-01

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

An overview of U.S. plant performance improvement through effective radioactive waste management

International Nuclear Information System (INIS)

Sieberling, R.; Lyons, P.W.

1986-01-01

Radioactive waste volumes in the United States commercial nuclear power industry have declined over the last five years after increasing steadily in the late 1970's. This decline is considered to be especially significant because major backfits/modifications and major component repair/replacement at U.S. nuclear power stations have been accomplished during this period. This paper analyzes the key reasons for this performance and outlines the positive effects that this trend has upon overall nuclear plant performance. In reviewing plant performance data, there appears to be a direct correlation between overall plant performance and radioactive waste performance. For example, plants with high capacity factors, strong industrial safety programs, low collective man-rem, and good contamination controls, generally have a history of low radioactive waste volumes. The reasons for this excellent performance is the result of direct management and supervisory involvement in plant operations, implementation of high standards of performance, and monitoring performance against these standards. Using the techniques outlined in this paper, the U.S. nuclear power industry can continue to reduce the volume of low-level solid radioactive waste being generated

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

International Nuclear Information System (INIS)

Ball, J.R.

1986-04-01

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

Application of advanced technology to LMR control

International Nuclear Information System (INIS)

Lindsay, R.W.

1989-01-01

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

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

International Nuclear Information System (INIS)

2013-01-01

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

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.

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

International Nuclear Information System (INIS)

1986-11-01

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

Big Rock Point: 35 years of electrical generation

International Nuclear Information System (INIS)

Petrosky, T.D.

1998-01-01

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

LBB application in Swedish BWR design

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

1987-02-01

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

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

International Nuclear Information System (INIS)

2013-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

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

Energy Technology Data Exchange (ETDEWEB)

West, C D [comp.

1990-05-01

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

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

International Nuclear Information System (INIS)

West, C.D.

1990-05-01

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

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Performance assessment of containment filtered venting system with Venturi scrubber

International Nuclear Information System (INIS)

Adinarayna, K.N.V.; Ali, Seik Mansoor; Balasubramaniyan, V.

2015-01-01

Venting through appropriate filtration systems is now being considered as a severe accident management strategy for maintaining the containment integrity and also as a means to reduce the radiological consequences to the public and environment. The option of filtered containment venting appears to have assumed significance in the post- Fukushima accident backdrop. Back-fitting of a suitable Venturi scrubber based CFVS for the Indian BWRs (TAPS- 1 and 2) at Tarapur is now being contemplated. Several key issues need to be carefully addressed for ensuring the desired functional capability of such a system. At the outset, this paper highlights a few thermal hydraulic issues that are of interest from regulatory perspective. This is followed by a detailed description of the mathematical models developed for assessing the depressurization characteristics of CFVS, energy absorption capacity of the Scrubber Tank (ST) water inventory, iodine removal and aerosol retention capability etc. Finally, application of these models to investigate the response of CFVS under twin unit SBO conditions in TAPS-1 and 2 is presented. The studies presented here give insight into the key variables affecting the CFVS performance and would be useful to both the system designer as well as the regulator. (author)

The safety of nuclear power plants in Eastern Europe

International Nuclear Information System (INIS)

Hoehn, J.; Niehaus, F.

1997-01-01

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

Application of advanced technology to LMR control

International Nuclear Information System (INIS)

Lindsay, R.W.

1989-01-01

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

Structural experiences at the Kewaunee Nuclear Power Plant

International Nuclear Information System (INIS)

Setlur, A.V.

1983-01-01

This paper discusses the original structural and geotechnical design and subsequent structural experience at the Kewaunee Nuclear Power Plant. The original design of the 535 MWe Westinghouse two loop PWR nuclear plant operated by Wisconsin Public Service Corporation, was started in 1967 and was completed in 1974 when the unit was put into commercial operation. Since 1974 a number of changes in the regulations and additional requirements have been imposed on operating reactors. The paper traces the influence of the original plant criteria on the backfit evaluations and the minimal physical changes required in the plant's structures and components to comply with the new requirements. In addition, the unique design features and construction challenges of the original design are discussed. Kewaunee Nuclear Power Plant has had one of the best operating performance records in the world. Also, the exposure to radiation for plant personnel and radioactive waste generation has been significantly lower than the average. This has been achieved by a conscientious team effort of all parties involved. Some of the more significant structural design features contributing to the excellent performance is detailed in this paper. (orig.)

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

International Nuclear Information System (INIS)

Raetzke, C.; Micklinghoff, M.

2006-01-01

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

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

International Nuclear Information System (INIS)

Woods, H.W.

1993-10-01

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

AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

International Nuclear Information System (INIS)

Hansen, W.

2003-01-01

Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de

Applicability of ASME sections III and VIII and of B31.1 and B31.3 to DOE facilities

International Nuclear Information System (INIS)

Antaki, G.A.

1993-01-01

DOE order 6430.1A Section 1300-3.2 requires that open-quotes....safety class items shall be designed to the ASME Boiler and Pressure Vessel Code (ASME Section III) or to other comparable safety-related codes and standards...close quotes. This requirement raises a host of technical and practical questions which, to the author's knowledge, have not been fully addressed in the past. This paper attempts to cover the following essential points, in order: Evolution of industry reference codes, Code scope, Safety margins, Logistical considerations, Costs, Backfit considerations. These points are covered in the context of a reference safety class piping and vessel system at a DOE facility which processes radioactive fluids, and which this paper calls the open-quotes reference DOE nuclear facilityclose quotes. In the conclusion, the author proposes three alternatives for code applicability which are ranked technically as open-quotes goodclose quotes, open-quotes closer to 6430.1Aclose quotes and open-quotes closest to 6430.1Aclose quotes. It is however questionable whether the alternatives which are labeled open-quotes closerclose quotes and open-quotes closestclose quotes are practically viable, as will be discussed

IGSC: using rocket science for increased power, 100% CO{sub 2} capture, and no NOx or SOx

Energy Technology Data Exchange (ETDEWEB)

Griffiths, J. [Jacobs Engineering, London (United Kingdom)

2009-02-15

The integrated gasification steam cycle (IGSC) is a new way of generating electricity from coal, which, when backfitted to an existing coal fired power plant, not only increases power output substantially but also allows 100% removal of carbon dioxide. The basic concept of IGSC is gasification of coal in a quench gasifier, followed by combustion of the resulting syngas, with oxygen and water, in a modified gas turbine fitted with a novel form of oxy-burner, derived from rocket technology, the CES burner. This turbine, which has no compressor and is called the 'fired expander', drives a generator. The exhaust from the combustion is passed through an HRSG and the resulting steam used to generate further electricity. Downstream of the HRSG, the exhaust gases, which consist of steam mixed with CO{sub 2}, are directly quenched with circulating cold water to condense all the stream, leaving the CO{sub 2} to be collected and compressed. It was decided to concentrate the main development and evaluation of the IGSC process on the retrofit case, focusing on the retrofit of 500 MWe power plants in the UK, with a consequent need for flexibility in operation. 12 figs., 1 tab.

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-01-15

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

Hydrogen removal from LWR containments by catalytic-coated thermal insulation elements (THINCAT)

International Nuclear Information System (INIS)

Fischer, K.; Broeckerhoff, P.; Ahlers, G.; Gustavsson, V.; Herranz, L.; Polo, J.; Dominguez, T.; Royl, P.

2003-01-01

In the THINCAT project, an alternative concept for hydrogen mitigation in a light water reactor (LWR) containment is being developed. Based on catalytic coated thermal insulation elements of the main coolant loop components, it could be considered either as an alternative to backfitting passive autocatalytic recombiner devices, or as a reinforcement of their preventive effect. The present paper summarises the results achieved at about project mid-term. Potential advantages of catalytic thermal insulation studied in the project are:-reduced risk of unintended ignition,;-no work space obstruction in the containment,;-no need for seismic qualification of additional equipment,;-improved start-up behaviour of recombination reaction. Efforts to develop a suitable catalytic layer resulted in the identification of a coating procedure that ensures high chemical reactivity and mechanical stability. Test samples for use in forthcoming experiments with this coating were produced. Models to predict the catalytic rates were developed, validated and applied in a safety analysis study. Results show that an overall hydrogen concentration reduction can be achieved which is comparable to the reduction obtained using conventional recombiners. Existing experimental information supports the argument of a reduced ignition risk

Basic national requirements for safe design, construction and operation

International Nuclear Information System (INIS)

Franzen, L.F.

1980-01-01

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

PSA Level 2 activities for RBMK reactors

International Nuclear Information System (INIS)

Gubler, R.

1998-01-01

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

Biblis 1,300 MW unit B in operation for thirty years

International Nuclear Information System (INIS)

Lauer, H.

2006-01-01

Over the past thirty years, unit B of the Biblis nuclear power station has contributed towards safe, reliable and non-polluting electricity generation. Unit B was the first 1,300 MW unit in the world confirming the good experience accumulated in the construction and operation of plants this size. It laid the foundation for the advanced development of nuclear power plants up to the convoy line. The good availability this plant has achieved in its operating cycles proves the mature state of technology and the high level of qualification and motivation of the power plant staff and the personnel of external firms contracted to work for Biblis. Biblis B has served as a reference plant in German nuclear safety research, demonstrating that units this size can be operated safely. Unit B is also living proof of the possibility to raise nuclear power plants built in the seventies to the current state of the art by implementing the appropriate backfitting and modernization measures. Today, Biblis B is operated at a safety level clearly higher than that to be achieved for new plants internationally. This is also evident from comparison with the guiding values for the safety of new facilities as published by the International Atomic Energy Agency (IAEA). (orig.)

Annual meeting on nuclear technology 2005. Proceedings

International Nuclear Information System (INIS)

2005-03-01

The proceedings of the annual meeting on nuclear technology 2005 covers the following issues: (1) reactor physics and methods of calculation: design and transients; method development and validation; (2): thermodynamics and fluid dynamics: analytical thermohydraulics for existing reactors; experiments and operational behavior; analytical methods for innovative reactors; (3) Safety of nuclear installations - methods, analysis, results: special problems; PSA and in-vessel phenomena; ex-vessel phenomena; (4) front end and back end of the fuel cycle, radioactive waste, storage: intermediate storage of fuel elements, waste treatment, (5) fuel elements and core components: fuel elements, new methods in the interpretation, manufacturing and service; (6) operation of nuclear installations: experience with the operation of NPPs; management systems, digital instrumentation and control of NPPs revision management; (7) decommissioning of nuclear installations: concepts and strategies for decommissioning and dismantling; experiences with decommissioning projects; (8) fusion technology: fusion facilities; materials and test facility; cryo technique and simulations; (9) research reactors: building new and backfitting of existing research reactors; current development; dismantling of research reactors; (10) advanced reactor concepts, energy systems, energy economics; (11) communication with the public; (12) component materials, fabrication and service behavior: degradation effects of component materials; component behavior; (13): radiation protection: PSA and in-vessel phenomena, ex-vessel phenomena.

Regulatory aspects of radiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Randall, P.N.

1979-01-01

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

Some practical implications of source term reassessment

International Nuclear Information System (INIS)

1988-03-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Containment venting sliding pressure venting process for PWR and BWR plants

International Nuclear Information System (INIS)

Eckardt, B.

1991-01-01

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

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

International Nuclear Information System (INIS)

Feuerbach, R.

2002-01-01

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

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

International Nuclear Information System (INIS)

Bekiarian, A.; Le Bastard, G.

1991-01-01

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

EPRI's zebra mussel monitoring and control guidelines

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

Cohen, S.; Mann, B.

1987-01-01

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

Papers: Congresses and Conferences, 2002-2003

International Nuclear Information System (INIS)

2004-01-01

Empresarios Agrupados (EA) is a leading architect-engineering organisation in Spain with significant international experience, providing a complete range of consulting, project management,engineering and design, procurement, construction, management, plant testing, safety assessment, quality assurance, as well as plant operation and maintenance support services to the electric utility industry. Founded in 1971, EA has a permanent multidisciplinary staff of approximately 1000, 65% of whom are university graduates, involved in engineering projects and services in the electric utility sector. Serving the electric utility industry is one of EA's primary objectives as an architect-engineering company. In the field of power generation, EA's work includes the design, construction and operation support of nuclear, fossil-fired and hydroelectric power plants and radioactive waste management facilities, as well as the safety assessment, modernisation, backfitting, re powering and life extension of operating plants and facilities. Services provided by EA in the field of power generation are: . Feasibility studies . Site selection and project development studies . Project management . Engineering and design . Procurement management . Construction management . Plant testing and startup . Plant operation and maintenance . Quality assurance/quality control EA has been the Architect-Engineer of power plant projects totalling more than 21,000 MWe of power generating capacity worldwide

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

International Nuclear Information System (INIS)

2011-12-01

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

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

International Nuclear Information System (INIS)

1998-01-01

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

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.

Seismic assessment of Kozloduy VVER 440, Model 230 nuclear power plant

International Nuclear Information System (INIS)

Monette, P.; Baltus, R.; Yanev, P.; Campbell, R.

1991-01-01

Excluding system design deficiency relative to US and Western Europe standards, it was found that the plant has many seismic vulnerabilities similar to those that existed in many of the US plants prior to about 1979 when the Systematic Evaluation Program was initiated. The primary coolant system has been substantially upgraded after the 1977 Vrancea earthquake. Other upgrades have been made to weak elements in the ECCS and electrical systems. There are still a number of components that could likely survive the currently defined Safe Shutdown Earthquake of 0.1 g but which would not meet current design standards. Many of the weakest components could be upgraded at a moderate cost to withstand a seismic event exceeding 0.1 g. Current studies of the site seismicity lean toward a higher peak ground acceleration and increased amplification of building motion, thus backfits that have been accomplished may become marginal for newly defined loads. However the proper consideration of soil structure interaction and detailed structural analysis using less conservative modeling assumptions, could mitigate the impact of increasing the seismic input and limit the amount of reinforcement required. In the interim, substantial improvements to seismic safety could be accomplished by simple, inexpensive modifications to equipment anchorage and some achievable improvements to connection detail of the precast concrete structures. (author)

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

International Nuclear Information System (INIS)

Ilg, Ulf

2008-01-01

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

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

International Nuclear Information System (INIS)

1998-01-01

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

Reactor risk reference document: Main report: Draft for comment

International Nuclear Information System (INIS)

1987-02-01

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

Upgrading instrumentation and control systems for plant safety and operation

International Nuclear Information System (INIS)

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

1997-01-01

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

Cost-benefit considerations in regulatory analysis

International Nuclear Information System (INIS)

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

1995-10-01

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

A study on items necessary to develop the requirements for the management of serious accidents postulated in fuel fabrication, enrichment and reprocessing facilities

International Nuclear Information System (INIS)

Takanashi, Mitsuhiro; Yamate, Kazuki; Asada, Kazuo; Yamada, Takashi; Endo, Shigeki

2013-05-01

The purpose of this study is to supply the points to discuss on new rules of fuel fabrication, enrichment and reprocessing facilities (hereinafter referred to as 'fuel cycle facilities') conducted by Nuclear Regulation Authority. Requirements for management of serious accidents in the fuel cycle facilities were summarized in this study. Taking into account the lessons learned from the accident of TEPCO Fukushima Daiichi Nuclear Power Plant in Mar. 2011, Act for the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors was amended in June 2012. The main items of the amendment were as follows: Preparation for the management of serious accidents, Introduction of evaluation system for safety improvement, Application of new standards to existing nuclear facilities (back-fitting). Japan Nuclear Energy Safety organization (JNES) conducted a fundamental study on serious accidents and their management in the fuel cycle facilities and made a report. In the report, the concept of Defense in Depth and the definition of serious accidents for the fuel cycle facilities were discussed. Those discussions were conducted by reference to new regulation rules (draft) for power reactors and from the view of features of the fuel cycle facilities. However, further detailed studies are necessary in order to clarify some issues in it. It was also reflected opinions from experts in JNES technical meetings on accident management of the fuel cycle facilities to brush up this report. (author)

Proceedings of an international workshop on historic dose experience and dose reduction (ALARA) at nuclear power plants

International Nuclear Information System (INIS)

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

1985-06-01

Dose reduction data and experience from 28 foreign and 10 US nuclear power plants was examined to determine causes for the wide variations in occupational dose from country to country. Major topics discussed were: steam generator and refueling maintenance problems; utility and supplier ALARA programs; effectiveness of dose-reduction modifications; attitudes and training; current and future dose-reduction research. While many parameters contribute to differences of occupational doses between plants from different nations, it is clear that most US plants have higher collective dose equivalent per reactor per megawatt-year than most other countries, even for plants of similar size and age. Worldwide, Finnish and Swedish plants, both PWR and BWR, have achieved the lowest values. Major factors which contribute to low doses include: (1) minimization of cobalt in primary system components exposed to water; (2) careful plant design, layout and component segregation and shielding; (3) plant standardization; (4) selection of components and systems for increased reliability; (5) management interest and commitment; (6) minimum number of workers and in-depth worker training; (7) careful control of primary system oxygen and pH; (8) good primary system water purity to minimize corrosion product formation; (9) use of special tools and robotics; (10) decontamination and passivation of primary systems and components; and (11) extent of backfitting and mandated inspections

Potential of low-temperature nuclear heat applications

International Nuclear Information System (INIS)

1986-12-01

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

Sensitivity analysis of the reactor safety study. Final report

International Nuclear Information System (INIS)

Parkinson, W.J.; Rasmussen, N.C.; Hinkle, W.D.

1979-01-01

The Reactor Safety Study (RSS) or Wash 1400 developed a methodology estimating the public risk from light water nuclear reactors. In order to give further insights into this study, a sensitivity analysis has been performed to determine the significant contributors to risk for both the PWR and BWR. The sensitivity to variation of the point values of the failure probabilities reported in the RSS was determined for the safety systems identified therein, as well as for many of the generic classes from which individual failures contributed to system failures. Increasing as well as decreasing point values were considered. An analysis of the sensitivity to increasing uncertainty in system failure probabilities was also performed. The sensitivity parameters chosen were release category probabilities, core melt probability, and the risk parameters of early fatalities, latent cancers and total property damage. The latter three are adequate for describing all public risks identified in the RSS. The results indicate reductions of public risk by less than a factor of two for factor reductions in system or generic failure probabilities as high as one hundred. There also appears to be more benefit in monitoring the most sensitive systems to verify adherence to RSS failure rates than to backfitting present reactors. The sensitivity analysis results do indicate, however, possible benefits in reducing human error rates

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Benjamin, A S; Kunsman, D M; Williams, D C [Sandia National Laboratories, Albuquerque, NM (United States); Boyd, G J; Lewis, S R [Safety and Reliability Optimization Services, Inc., Knoxville, TN (United States); Smith, L N [Science Applications International Corporation, Albuquerque, NM (United States)

1987-04-01

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

NUMARC view of license renewal criteria

International Nuclear Information System (INIS)

Edwards, D.W.

1989-01-01

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

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

International Nuclear Information System (INIS)

Lindauer, E.

2002-01-01

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

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

Science.gov (United States)

Ma, Yanyuan; Wang, Yuanjia

2014-04-15

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

A human reliability assessment screening method for the NRU upgrade project

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Impact of extreme load requirements and quality assurance on nuclear power plant costs

International Nuclear Information System (INIS)

Stevenson, J.D.

1993-01-01

Definitive costs, applicable to nuclear power plant concrete structures, as a function of National Regulatory Requirements, standardization, the effect of extreme load design associated with both design basis accidents and extreme external events and quality assurance are difficult to develop since such effects are interrelated and not only differ widely from country to country, project to project but also vary in time. Table 1 shows an estimate of the of the overall plant cost effects of external event extreme load design on nuclear power plant design for the U.S -and selected foreign countries for which experience with LWRs exist- Germany is the most expensive primarily due to a military aircraft crash resistance. However, the German requirement for 4 safeguards trains rather than 2 and the containment design requirement to consider one Steam Generator blowdown concurrent with a RCS blowdown. This presentation will concentrate on the direct current impact extreme load design and quality assurance have on concrete structures, systems and components for nuclear plants. This presentation is considered timely due to the increased interest in the c potential backfit of Eastern European nuclear power stations of the WWER 440 and WWER 1000 types which typically did not consider the extreme loads identified in Table 1 and accident loads in Table 3 and quality assurance in Table 5 in their original design. Concrete structures in particular are highlighted because they typically form the last barrier to radioactive release from the containment and other Safety Related Structures

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

International Nuclear Information System (INIS)

Hashizume, Hidetoshi; Aoki, Takayuki

2015-01-01

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

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

International Nuclear Information System (INIS)

2011-01-01

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.)

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

International Nuclear Information System (INIS)

1994-01-01

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

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

International Nuclear Information System (INIS)

Becker, Oda; Lorenz, Patricia

2013-04-01

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

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.

Reduction of PWR containment pressure after hypothetical accidents by water-cooling of the outer containment surface - annular space spray system

International Nuclear Information System (INIS)

Cremer, J.; Dietrich, D.P.; Roedder, P.

1980-12-01

The consequences of a core melt-out accident in the vicinity of a nuclear power station are determined by the integrity of the safety containment. This can be adversely affected by different events during the course of the core melt-out accident. The most important phenomenon is the contact between the melt and sump water. Due to the evaporation of the sump water, there is a continuous rise in pressure of the safety containment, which finally leads to failure due to excess pressure. In order to reduce the fission product release due to the resulting leakage, one must try to reduce the pressure as quickly as possible. As heat cannot be removed from the steel containment to the environment because of the thick concrete containment, it is best to bypass the insulating effect of the concrete by cooling the steel containment from outside. The aim of this investigation is therefore to work out a technically relatively simple system, which offers the possibility of backfitting, setting to work and repair. Such a system is an annular space spray system, by which the annular space between the concrete and steel containment has water pumped to the level of the dome and evenly sprayed over the top hemisphere. Mobile pumps on fire engines belonging to the fire brigade are sufficient to supply the cooling water and these will be available some hours after the accident occurs. The used spray water without any radioactive components is collected outside the reactor building and/or drained off. (orig./GL) [de

Atmospheric dispersion models for real-time application in the decision support system being developed within the CEC

International Nuclear Information System (INIS)

Mikkelsen, T.

1992-01-01

A number of Commission of the European Communities member states are presently coordinating their research and development of a ''Real-time On-line DecisiOn Support'' (RODOS) for emergency assistance in the event of nuclear accident. In addition to atmospheric dispersion, the system involves multiple other radiological disciplines. The ability to estimate a specific atmospheric dispersion scenarios in real-time becomes a first-priority task and is of uttermost importance for the subsequent success or failure of such a comprehensive decision support system to guide off-site emergency situations. No single dispersion model is at present able to cover all possible release-types and scales of dispersion. A hierarchy of atmospheric flow and dispersion models is presently being ranked for suitability to real-time calculate air and integrated-air concentrations. Starting at the short-range scale, models are discriminated with respect to principle, adequacy and flexibility, CPU-time constraints, experimental evaluation record, instantaneous or short-time release handling, deposition measures (wet and dry), input and output data flexibility and uncertainty-handling and model-interpretation. Additional features of particular importance are: Robustness in schemes for meteorological preprocessing of meteorological input data, and on-line backfitting and data-assimilation procedures. Models demonstrating practical and operational use, including real-time operational experience, have in this context the highest priority, as opposed to the more sophisticated and theoretical ''development-type'' models. Real-time methods founded on our present knowledge and data concerning flow and dispersion in the atmospheric boundary layer, are of primary interest. (au) (18 refs.)

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

Kroeger, W.

1992-09-01

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

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

International Nuclear Information System (INIS)

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

1997-09-01

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

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

International Nuclear Information System (INIS)

Raetzke, C.; Micklinghoff, M.

2005-01-01

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

An analysis of nuclear power plant operating costs

International Nuclear Information System (INIS)

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Benjamin, A S; Griesmeyer, J M; Haskin, F E; Kunsman, D M [Sandia National Laboratories, Albuquerque, NM (United States); Boyd, G J; Lewis, S R [Safety and Reliability Optimization Services, Inc., Knoxville, TN (United States); Helton, J C [Arizona State University, Tempe, AZ (United States); Smith, L N [Science Applications International Corporation, Albuquerque, NM (United States)

1987-04-01

The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a boiling water reactor with a Mark 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)

Regulatory aspects of NPP safety

International Nuclear Information System (INIS)

Stuller, J.; Brandejs, P.; Miasnikov, A.; Svab, M.

1999-01-01

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)

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

International Nuclear Information System (INIS)

Johanson, G.; Kreuser, A.; Pyy, P.; Rasmuson, D.; Werner, W.

2006-01-01

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

Safeguarding the functioning of I and C systems

International Nuclear Information System (INIS)

Koehler, M.; Schoerner, O.

1997-01-01

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

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

International Nuclear Information System (INIS)

1987-01-01

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

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

International Nuclear Information System (INIS)

Schubert, G.; Krause, H.

1993-12-01

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

The power of British Energy

International Nuclear Information System (INIS)

Hawley, R.

1997-01-01

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

Advanced human-system interface design review guidelines

International Nuclear Information System (INIS)

O'Hara, J.M.

1990-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-09-01

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

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

International Nuclear Information System (INIS)

2000-09-01

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

International status of application of probabilistic risk analysis

International Nuclear Information System (INIS)

Cullingford, M.C.

1984-01-01

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

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

International Nuclear Information System (INIS)

James, D.W.

1985-01-01

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

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

International Nuclear Information System (INIS)

Berger, E.; Tinic, S.

1988-01-01

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

Confirmation of soil radiation damping from test versus analysis

International Nuclear Information System (INIS)

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

1987-01-01

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

MOX fuel design and development consideration

International Nuclear Information System (INIS)

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

1997-01-01

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

Implementation of utilities operation and maintenance experience into the European pressurized water reactor design

International Nuclear Information System (INIS)

Zaiss, W.; Lallier, M.

1999-01-01

Since 1992 Electricite de France EDF and German Utilities GU work together with Nuclear Power International NPI, a subsidiary of Framatome and Siemens, in the development of the future European Pressurized Water Reactor EPR. The EPR is an evolutionary concept, based on the French N4 plants and the German KONVOI plants. From the beginning, experienced operation and maintenance people from the precursor plants participate at the design process. Their experience will lead to a plant, which is not only characterised by low investment costs, but also by good operability, high availability and low operation and maintenance costs. No expensive back-fittings should be necessary after commissioning, to reach these availability and maintenance targets. The utility specialists give design requirements for outage performance, system design, and layout. These design requirements are really determining the system performances, and not what was design basis before. It does not necessarily lead to system increases. Mainly it is a shifting of the emphasis to other items. There are even cases, where the system performances can be reduced. Mostly very small modifications, which are nearly cost neutral when implemented early in the design, have big impact on the further operation. If there are big cost influences, a sound balance between investment and gained availability is made together with the designers. There is very fruitful discussion between designers and operators, which is highly estimated by both sides. In this frame also new, revolutionary ideas are coming up, which are going mostly in the direction of investment cost reduction, without loosing operation freedom. It is the first time in Europe, that designers and operators are working so close together. It is also the first time, that the management and the decision making is dominated by the utilities. (author)

Designing for nuclear power plant maintainability and operability

International Nuclear Information System (INIS)

Pedersen, T.J.

1998-01-01

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

Plant Betterment as Anticipated Measure For Plant Life Management

International Nuclear Information System (INIS)

Louvat, J. P.

1991-01-01

A lot of modifications have been made since critically on each of the 28 standardized 900 MW class PWR units in France. Most of this technical upgrading was accomplished to facilitate operation, improve availability, or bring the unit design in line with evolving regulatory requirements, but a substantial part of the modifications was dedicated to Plant Life Management. As part of the program launched by EDF for plant life management, this paper introduces the Frustum's contribution for plant betterment and enhancement of reactor operation concurrently to ensure or extend plant service life. The solutions contemplated in this field are provided to reduce the frequency of unexpected reactor trip occurrences, to mitigate their negative effects or to smooth off the reactor operation and thus the magnitude of associated transients. The lifetime evaluation of NPP is basically an economical exercise, which tries to determine how long the operation of the plant will remain competitive, taking into account the long term perspective maintenance costs. There cannot be any conflict between lifetime and safety considerations, based upon the pituitary requisite that the safety requirement must be met at any time of the operation. Plant life management needs a consistent approach that can not be improvised on a case by case basis. Instead, it must be kept in mind from the very beginning of unit operation. This is the sense of the backfitting and technical upgrading carried out in France for the PWRs of the 900 MW class. It is thanks to this necessary anticipation that plant life will be actually managed, giving benefit both from the standpoint of availability and from that of the service lives of sensitive components. Substantial savings will thus be obtained

The Spanish nuclear adventure. From the past into the future

International Nuclear Information System (INIS)

Gonzalez de Ubieta, A.

1986-10-01

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

Comparison and lessons learned from plant specific PSA of German NPP

International Nuclear Information System (INIS)

Balfanz, Hans-Peter; Berg, H.P.

2000-01-01

PSA are launched in frame of Periodic Safety Reviews (PSR) in Germany. The aims are to identify overall safety level and relative weak points. Some backfitting measures have been realized for older plants to remove relative weak points and to bring these plants to the state of the art. In this field PSA is well accepted today and is seen as a valuable tool supplementing the deterministic analysis. Main application of PSA within PSR is planned to become mandatory as part of the revision of the German Atomic Energy Act. According to the German PSA Guideline plant specific PSA level 1+ were performed for all 19 In comparison with international practice German PSA are very detailed. Otherwise they do not handle all external events, non-power states and accident management measures as discussed before. The New PSA guideline will cover these aspects and therefore analysts have to take them into account in further PSA. Moreover gathering of plant specific data is needed. The development in this field is driven by the utilities (for instance in frame of their so-called ZEDB project). Public discussion about quantitative risk of industrial hazards is quite limited in Germany and PSA results have only few impacts to this respect. Independent from this PSA for NPP is understood as a diverse tool in supporting the deterministic licensing and supervision process. Risk based decision making as well as informed regulation are just only of the beginning. State of PSA of NPP in Germany, comparison of PSA result of different NPP, German PSA guideline and state of discussion of further development and recommendation of further development of PSA of NPP are discussed in this paper in more detail. (S.Y.)

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

International Nuclear Information System (INIS)

Smith, B. L.

2002-03-01

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

Comparison and lessons learned from plant specific PSA of German NPP

Energy Technology Data Exchange (ETDEWEB)

Balfanz, Hans-Peter [TUEV Nord, Hamburg (Germany); Berg, H.P.

2000-07-01

PSA are launched in frame of Periodic Safety Reviews (PSR) in Germany. The aims are to identify overall safety level and relative weak points. Some backfitting measures have been realized for older plants to remove relative weak points and to bring these plants to the state of the art. In this field PSA is well accepted today and is seen as a valuable tool supplementing the deterministic analysis. Main application of PSA within PSR is planned to become mandatory as part of the revision of the German Atomic Energy Act. According to the German PSA Guideline plant specific PSA level 1+ were performed for all 19 In comparison with international practice German PSA are very detailed. Otherwise they do not handle all external events, non-power states and accident management measures as discussed before. The New PSA guideline will cover these aspects and therefore analysts have to take them into account in further PSA. Moreover gathering of plant specific data is needed. The development in this field is driven by the utilities (for instance in frame of their so-called ZEDB project). Public discussion about quantitative risk of industrial hazards is quite limited in Germany and PSA results have only few impacts to this respect. Independent from this PSA for NPP is understood as a diverse tool in supporting the deterministic licensing and supervision process. Risk based decision making as well as informed regulation are just only of the beginning. State of PSA of NPP in Germany, comparison of PSA result of different NPP, German PSA guideline and state of discussion of further development and recommendation of further development of PSA of NPP are discussed in this paper in more detail. (S.Y.)

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

International Nuclear Information System (INIS)

Manners, S.; Blocquel, Ch.

1999-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Manners, S.; Blocquel, Ch

1999-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Berger, E [Basler and Hofmann AG, Consulting Engineers, Zurich (Switzerland); Tinic, S [Nordostschweizerische Kraftwerke AG, Baden (Switzerland)

1988-07-01

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

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.

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

International Nuclear Information System (INIS)

Degenhart, Christoph

2010-01-01

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

PSA in America

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

Schoenenberger, M.

2005-01-01

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

Absorber materials in CANDU PHWR's

International Nuclear Information System (INIS)

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

1995-03-01

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

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

International Nuclear Information System (INIS)

Gallucci, Raymond H.V.

2016-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Mitigation of release of volatile iodine species during severe reactor accidents - a novel reliable process of safety technology

International Nuclear Information System (INIS)

Guentay, S.; Bruchertseifer, H.

2010-01-01

In severe accidents, a significant risk for public health may be generated as a result of release of the gaseous iodine species into the environment through the containment leaks or containment venting filter systems with low retention efficiency. The elemental iodine and volatile organic iodides are the main gaseous iodine species in the containment. Potential release of large quantities of gaseous elemental iodine from the reactor coolant system or its radiolytic generation in the containment sump constitute the key source of gaseous elemental iodine in containment atmosphere. Iodine paint reactions as well as the reaction of iodine with organic residuals in sump water are the main mechanisms for the generation of high volatile organic iodides in the containment. Although very much desired, significant research activities conducted in 70's unfortunately did not create any technically feasible solution to mitigate iodine release into the environment under prevailing conditions. Development of a process leading to a fast, comprehensive and reliable retention of volatile iodine species in aqueous solution with an aim to implement for the severe accident management applications has been subject of a research project in the recent years at Paul Scherrer Institut. The process developed utilizes simultaneous use of two customary technical chemical additives in an aqueous solution. The results of the experimental program have demonstrated a fast and reliable destruction of high volatile organic iodine species and fast reduction of elemental iodine into iodide ions in aqueous solutions and an efficient mitigation of the re-formation of gaseous iodine from iodide ions. Investigations covered a broad range of anticipated severe accident conditions in the containment. The project additionally focused on possible application of the process to existing containment venting filter systems, specifically as a passive add-on for back-fitting. This paper describes the process

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

International Nuclear Information System (INIS)

Brey, Larry H.

2014-01-01

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

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.

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

International Nuclear Information System (INIS)

Murani, J.

1995-01-01

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

Cost effective decommissioning and dismantling of nuclear power plants

International Nuclear Information System (INIS)

Wasinger, Karl

2012-01-01

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

Development of the methodology for application of revised source term to operating nuclear power plants in Korea

International Nuclear Information System (INIS)

Kang, M.S.; Kang, P.; Kang, C.S.; Moon, J.H.

2004-01-01

Considering the current trend in applying the revised source term proposed by NUREG-1465 to the nuclear power plants in the U.S., it is expected that the revised source term will be applied to the Korean operating nuclear power plants in the near future, even though the exact time can not be estimated. To meet the future technical demands, it is necessary to prepare the technical system including the related regulatory requirements in advance. In this research, therefore, it is intended to develop the methodology to apply the revised source term to operating nuclear power plants in Korea. Several principles were established to develop the application methodologies. First, it is not necessary to modify the existing regulations about source term (i.e., any back-fitting to operating nuclear plants is not necessary). Second, if the pertinent margin of safety is guaranteed, the revised source term suggested by NUREG-1465 may be useful to full application. Finally, a part of revised source term could be selected to application based on the technical feasibility. As the results of this research, several methodologies to apply the revised source term to the Korean operating nuclear power plants have been developed, which include: 1) the selective (or limited) application to use only some of all the characteristics of the revised source term, such as release timing of fission products and chemical form of radio-iodine and 2) the full application to use all the characteristics of the revised source term. The developed methodologies are actually applied to Ulchin 9 and 4 units and their application feasibilities are reviewed. The results of this research are used as either a manual in establishing the plan and the procedure for applying the revised source term to the domestic nuclear plant from the utility's viewpoint; or a technical basis of revising the related regulations from the regulatory body's viewpoint. The application of revised source term to operating nuclear

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

International Nuclear Information System (INIS)

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

2013-03-01

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

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.

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

International Nuclear Information System (INIS)

2011-01-01

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

Cost effective decommissioning and dismantling of nuclear power plants; Kosteneffizienz bei Stilllegung und Rueckbau von Kernkraftwerken

Energy Technology Data Exchange (ETDEWEB)

Wasinger, Karl [AREVA NP GmbH, Offenbach (Germany)

2012-10-15

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

US nuclear safety review and experience

International Nuclear Information System (INIS)

Gilinsky, V.

1977-01-01

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

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

International Nuclear Information System (INIS)

Vakulovski, S.M.; Shershakov, V.M.; Borodin, R.V.; Vozzhennikov, O.I.; gaziev, Y.I.; Kosykh, V.S.; Makhon'ko, K.P.; Chumichev, V.B.

1994-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-03-15

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

Lessons Learned from the Fukushima Daiichi Accident, Actions Taken and Challenges Ahead

International Nuclear Information System (INIS)

Shimizu, Y.

2016-01-01

On 19 September, 2012, the Nuclear Regulation Authority (NRA) was established in light of lessons learned from the Fukushima Daiichi accident of 11 March 2011, to ensure that such accidents never happen again, to restore public trust in regulator both in Japan and abroad and to rebuild and foster a genuine safety culture by placing the highest priority on public safety. The NRA, an independent administrative commission of the Ministry of the Environment, is organized to separate the regulatory functions from the promotional functions of the use of nuclear energy within the government, and to independently implement its duties from the perspectives of neutrality and fairness based on its expertise. Having learned the lessons from the Fukushima Daiichi accident and with reference to IAEA safety standards, since its establishment, the NRA has endeavored to strengthen the regulatory requirements, in particular, for hazards such as tsunamis and earthquakes which may lead to common cause failures, and countermeasures against severe accidents. Under the new regulatory scheme, a back-fitting system was introduced. Emergency preparedness and response measures for nuclear facilities were also enhanced. As of end of March 2016, five reactors received NRA’s permission for changing their reactor installations based on the new regulatory requirements, and two nuclear power reactors have restarted their operations. In January 2016, at the request of Japan, the IAEA sent the IRRS mission team to Japan to assess the regulatory framework for nuclear and radiation safety. Through the self-assessment prior to the mission, the NRA has developed 22 action plans, including a) improvement of regulatory inspection, b) capacity building, and c) strengthening of safety research capability. The mission team has found that Japan’s nuclear regulator has demonstrated independence and transparency since it was set up in 2012. The team also noted that the NRA needs to improve the inspection

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

International Nuclear Information System (INIS)

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

2015-02-01

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

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

International Nuclear Information System (INIS)

2016-07-01

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

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

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

International Nuclear Information System (INIS)

Leidinger, Tobias

2013-01-01

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

ACR-1000: Operator - based development

International Nuclear Information System (INIS)

Shalaby, B.; Alizadeh, A.

2007-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2016-07-15

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

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

Debris impact on emergency coolant recirculation - summary and conclusions

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Zivelonghi, Alessandro

2011-01-01

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

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

International Nuclear Information System (INIS)

Erve, M.

1998-01-01

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

Hardware replacements and software tools for digital control computers

International Nuclear Information System (INIS)

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

1996-01-01

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

Solution for backfitting of a controlled atmosphere area in radiopharmaceuticals unit: the mobile unit of decontamination; Solution pour la mise en conformite d'une zone d'atmosphere controlee en unite de radiopharmacie: l'unite mobile de decontamination

Energy Technology Data Exchange (ETDEWEB)

Moreau, H.; Desruet, M.D. [Radiopharmacie, CHU de Grenoble, (France); Gallazzini-Crepin, C.; Calizzano, A.; Bourre, J.C. [clinique universitaire de medecine nucleaire, CHU de Grenoble, (France); Skalli, S.; Foroni, L. [pole de pharmacie, CHU de Grenoble, (France); Fagret, D. [clinique universitaire de medecine nucleaire, CHU de Grenoble, (France)

2009-05-15

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

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

International Nuclear Information System (INIS)

Martin, M.

2000-01-01

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

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

International Nuclear Information System (INIS)

Lafaille, J.P.

1993-01-01

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

Conclusions Drawn from the Investigation of LOCA-Induced Insulation Debris Generation and its Impact on Emergency Core Cooling (ECC) at German NPPs - Approach Taken by / Perspective of The German TSO (TuV)

International Nuclear Information System (INIS)

Huber, J.

2004-01-01

Initiated by the Barsebaeck incident in 1992 and the following activities related to the LOCA-induced insulation debris generation and its impact on emergency core cooling, investigations on German PWRs and BWRs were performed in these areas. The investigations on the German BWRs were carried out in detail immediately after the Barsebaeck incident in the years 1992 through 1994. Detailed investigations on the German PWRs started after the issue of the OECD report in 1996. Therefore the investigations on the impact of LOCA-induced insulation debris generation on strainer plugging carried out in Germany in the last years were focused mainly on the German PWRs. In the framework of these investigations of the German PWRs, the relevant parameters and phenomena were investigated in detail by the plant owners in the years from 1997 through 1999. The results were summarised in reports for each plant. The main results of the investigations conducted by the plant owners were that the plant owners considered backfitting in German PWRs is not necessary to guarantee emergency core cooling following a LOCA with insulation debris generation. As the technical support organisation for the German Bavarian and Hessian state authority, the TUV Suddeutschland was called upon to examine these investigations and the conclusions drawn by the plant owners. We compared each of the parameters and phenomena against the state of knowledge. The results of our examination in 1999 showed that the investigations of the plant owners were generally correct, but we stated also, that due to existing uncertainties, further investigations are necessary to validate the results. To meet these demands, the plant owners installed a working group for planning and performing newer, more realistic large-scale experiments (scaling factor 1:4) to investigate the transport mechanism of the insulation material within the containment sump, the head loss at the strainers and to estimate the amount of insulation

Experience and trends at the Belgonucleaire plant

International Nuclear Information System (INIS)

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

2000-01-01

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

Aging evaluation methodology of periodic safety review in Korea

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

2001-01-01

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

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.)

ILK statement on determining operation periods for nuclear power plants in Germany

International Nuclear Information System (INIS)

2005-09-01

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.)

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

Energy Technology Data Exchange (ETDEWEB)

Bittermann, Dietmar; Eckardt, Bernd A.; Lechleuthner, Michael [Framatome ANP GmbH, Erlangen (Germany)

2003-04-01

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 4{sup th} 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 H{sub 2}-control, filtered venting, core retention devices and atmosphere sampling. The following systems are introduced for operating as well as for future plants: {center_dot} 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. {center_dot} In situ Post Accident Sampling System (In situ-PASS) are introduced for the purpose of

New system technologies implemented at Kozloduy 3 and 4 (WWER 440-230) for containment leakage and H2 control in severe accident situations - Design, qualification, installation, commissioning

International Nuclear Information System (INIS)

Feuerbach, R.; Eckardt, B.; Kastner, B.

2005-01-01

In order to reduce the residual risk associated with hypothetical severe nuclear accidents, systems and components for filtered containment venting and H 2 reduction were developed. During severe accident scenarios large quantities of hydrogen and radioactive material may be released into the containment atmosphere within a short period of time. In the event of internal over pressurization due to hypothetical severe accident sequences a pressure barrier system has to be created to confine the activity in the containment. Unavoidable releases of activity to the environment have to be minimized to a great extent as possible. Research into the hypothetical event of core melt accidents has continued and new accident mitigation technologies have been developed. Decisions have been taken to implement these new mitigation measures in operating nuclear power plants to mitigate severe accidents consequences. In order to prevent loss of containment integrity as a result of over pressurization, nuclear power plants in the Federal Republic of Germany as well as in most other European countries have been or will be back-fitted with systems for filtered venting of the containment atmosphere and systems for H 2 -control. Similar technologies for containment venting system and H 2 control have been now implemented in the first WWER 440-230 units of Kozloduy 3 and 4. Following OECD recommendations sever accident situations were analyzed and a design of countermeasures have been performed. Main goal of the developed countermeasures was to overcome the WWER 440-230 containment design specifics like, leakage rate behavior, limited available containment volume combined with the feature of high availability of electrical supply at multiple plant sites. Further more the design of counter measures considers the common use for Kozloduy unit 3 and 4. The analysis of postulated severs accident situation - without countermeasures - showed significant increase of H 2 /O 2 concentration in the

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

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)

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)

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

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

International Nuclear Information System (INIS)

2012-12-01

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

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

International Nuclear Information System (INIS)

2011-01-01

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

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

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

International Nuclear Information System (INIS)

2010-07-01

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 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 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 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 in 2005

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

International Nuclear Information System (INIS)

2007-07-01

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

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

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

International Nuclear Information System (INIS)

Calhoun, David; Shepherd, Stephen

2001-01-01

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

Occupational exposure at pressurised water reactors: international intercomparison from 1975-1989

International Nuclear Information System (INIS)

Bendittini, M.; Tabare, M.

1991-11-01

generation of reactors than that of the older. - The impact of 'operation-maintenance feedback'. An annual increase of 2% can be seen for the collective exposure of the oldest generation while values decrease slowly for the newest one. - There is no significant effect of the electrical output on occupational exposure for a same g generation. - The increase of the mean duration of cycle for post 1980 units has a positive impact on the mean collective dose per unit. - Oldest units had to support the effect of backfitting modifications needed for reactor safety. - The distribution of annual collective dose values is smaller for units of the last generation than for those of the older. All these remarks show the positive impact of feedback showing the evolution of mean collective dose per unit or per GWh assessed for each country according to the calendar years. Indicators show a convergence of values since 1982. The cumulative collective dose and electrical production since 1975 per country is shown. Most curves present an inflexion towards a lower dose for the latest TWhs produced. However, Swiss, French and Finnish figures show a more stable situation since the beginning of operation. As far as utility personnel and contractors are concerned, their respective contributions to the collective exposure are quite different (about 70% for the later). The outside worker share nevertheless varies between countries, from less than 50% in Switzerland to more than 90% in Spain and in Japan. The evolution of mean annual individual doses for utility workers (all countries, except Spain, Sweden and the US) is shown. For contractors who are mostly transient workers, the individual exposure is much more difficult to estimate. In each plant, fraction of individual doses associated to maintenance jobs being performed during outage periods are measured, and thus the distribution of individual doses yearly received at the level of the plant can be provided. However, in the absence of a

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Financing long term liabilities (Germany)

International Nuclear Information System (INIS)

2003-01-01

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

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

International Nuclear Information System (INIS)

Miller, Don W.

2001-01-01

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

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

International Nuclear Information System (INIS)

2001-01-01

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

Fuel for new Russian reactor VVER-1200

Energy Technology Data Exchange (ETDEWEB)

Vasilchenko, Ivan Nikitovich [GRPress, 21, Ordzhonikidze Street, 142103 Podolsk, Moscow region (Russian Federation)

2009-06-15

guiding channels and spacing grids. The spacing grid is made of the honeycomb cells welded to each other by resistance welding. The grid height is increased to prevent warping under thermomechanical influence of fuel rod bundle. Due to improvement in design of FA top and bottom parts the height of fuel rods and, accordingly, of the core is increased. The fuel assembly contains the easily removed top nozzle, joined to the guiding channels. The fuel rod end pieces are installed into the lower steel grid. The number and structure of grids assure absence of fuel rod fretting wear. Stream-lined and rigid structure of grids ensures a possibility of performing the handling procedures at increased rate. As, for instance, the core loading and unloading can be performed at the rate to 4 m/min that makes reduction in the reactor refueling time and increase in load factor. The alloy E-110, the same as in the prototype, is used as the fuel rod cladding material. Its high corrosion resistance is known also at increased parameters of new reactor. To improve the operational reliability of assemblies the design of anti-debris filter is developed. Results of FA operation show that there is not only a geometrical stability of the structure, but also a high residual life. The same is referred to fuel rods as well. All these factors made it possible to start implementation of the program of operating Units power increase and transition to longer fuel cycles at Russian NPPs with such type of reactors. A complete set of TVS-2M is under fabrication for the first loading of Unit 2, Rostov NPP, to be commissioned. Increase in the core height required modernization of ICIS. This experience makes it possible to use such a structure for AES-2006 with a back-fit. The attractive feature of TVS-2M type structure is its ease of manufacture, a high degree of automation in manufacturing. This will provide for not only maintaining a high quality of fuel but also a possibility of deliveries for demands