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1.
Hui-Wen Huang Chunkuan Shih Swu Yih Ming-Huei Chen Jiin-Ming Lin 《Nuclear Engineering and Design》2007,237(9):955-971
One of the major concerns when employing digital I&C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I&C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes “simulator/plant model testing” technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I&C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran–ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then futher be applied to hazard analysis for operator/digital I&C interface interaction failure study, and the hardware-in-the-loop fault injection study. 相似文献
2.
Aging degradation in nuclear power plants must be controlled to prevent safety margins from declining below limits provided in plant design bases. The NPAR Program and other aging-related programs conducted under the auspices of the NRC Office of Research are developing needed technical guidance for control of aging. Results from these programs, together with relevant information developed by industry and elsewhere, are implemented through various ongoing NRC and industry programs and initiatives as well as by means of conventional regulatory instruments. The aging control process central to these efforts consists of three key elements: (1) selection of components, systems, and structures (CSS) in which aging must be controlled, (2) understanding of the mechanisms and rates of degradation in these CSS, and (3) managing degradation through effective surveillance and maintenance. These elements are addressed in Recommended Practices Guidance that integrates information developed under NPAR and other studies of aging into a systems-oriented format that tracks directly with the Safety Analysis Reports and with the NRC Standard Review Plan (NUREG-0800). 相似文献
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生态环境部第8号令《核动力厂、研究堆和核燃料循环设施安全许可程序规定》对核动力厂、研究堆和核燃料循环设施运行许可证件延续事项作出了新的规定。为推动我国研究堆老化管理标准体系建立,分析了我国研究堆延寿审查策略发展历程,结合高通量工程试验堆等研究堆运行许可证有效期延续申请审查工作中的几个关键问题,提出了以定期安全审查为主、重点依据老化管理并兼顾技术规格书审查及差异性审查的审查策略,研究成果为我国研究堆老化管理法规标准的建立提供了实践经验及理论指导依据。 相似文献
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The large scale seismic test (LSST) program at Hualien, Taiwan, is a follow-on to the soil-structure interaction (SSI) experiments at Lotung, Taiwan. The planned SSI studies are performed at a stiff soil site in Hualien, Taiwan, that historically has had slightly more destructive earthquakes in the past than Lotung. The objectives of the LSST program are as follows: to obtain earthquake-induced SSI data at a stiff soil site having similar prototypical nuclear power plant soil conditions; to confirm the findings and methodologies validated against the Lotung soft soil SSI data for prototypical plant condition applications; to validate further the technical basis of realistic SSI analysis approaches; to support further the resolution of USI A-40 “Seismic Design Criteria” issue. These objectives are accomplished through an integrated and carefully planned experimental program consisting of soil characterization, test model design and field construction, instrumentation layout and deployment, in situ geophysical information collection, forced vibration test, and synthesis of results and findings. The Hualien LSST is a joint effort among many interested parties. The Electric Power Research Institute (EPRI) and Taiwan Power Company (Taipower) are the organizers of the program and have the lead in planning and managing the program. Other organizations participating in the LSST program are the US Nuclear Regulatory Commission (NRC), Central Research Institute of Electric Power Industry (CRIEPI), Tokyo Electric Power Compamy (TEPCO), Commissariat A L'Energie Atomique (CEA), Electricité de France (EdF), Framatome, Korea Electric Power Corporation (KEPCO), Korea Institute of Nuclear Safety (KINS), and Korea Power Engineering Company (KOPEC). The LSST array started operation in June 1993, and is envisioned to be of five years duration. 相似文献
6.
D. Guzy 《Nuclear Engineering and Design》1987,98(2)
The USNRC Piping Review Committee (PRC) was formed in 1983 with a charter to review NRC piping criteria, to recommend changes to this criteria, and to identify areas that would benefit from future research. This overview will outline the NRC-sponsored research being conducted to address those PRC recommendations concerning the design of nuclear piping systems to withstand dynamic loads. A key element of this research is the joint EPRI/NRC “Piping and Fitting Reliability Research Program.” This program consists of dynamic capacity testing of piping at the system, component, and specimen levels, plus analyses needed to support recommendations for changes to the ASME Code. As part of NRC's contribution to the EPRI/NRC program, a pipe system capacity test will be conducted at ETEC. The “Nonlinear Piping Response Prediction” project at HEDL is evaluating nonlinear response prediction techniques with differing degrees of complexity and will compare the various analytical results both with each other and with physical benchmarks such as the ETEC test. An ORNL project is developing nozzle design guidance that will provide a more realistic basis for evaluating the higher nozzle loads that will result from the more flexible piping systems design that are being considered. INEL will evaluate high frequency damping by considering the existing high frequency data and by conducting high frequency/high stress tests on two piping systems. LLNL is now conducting studies to more completely assess the uncertainties in the seismic response of building structures and piping systems. As a follow-on to the research efforts reported in NUREG/CR-3811, BNL will conduct additional studies to improve combinational procedures for piping response spectra analyses. 相似文献
7.
A design for an innovative, passively safe 10 MWe power plant based on the proven pressurized water reactor technology has been developed. The plant incorporates an innovative design approach to achieve “walk-away” safety and includes significant simplification and elimination of systems and components when compared to the current generation commercial nuclear power plants. The plant has been designed such that the majority of the equipment will be pre-assembled as modules at off-site facilities and shipped to the site on trucks for installation. This approach will provide shorter construction schedules and improved quality control. 相似文献
8.
S. Nisan G. Caruso J. -R. Humphries G. Mini A. Naviglio B. Bielak O. Asuar Alonso N. Martins L. Volpi 《Nuclear Engineering and Design》2003,221(1-3):251
This paper summarises our recent investigations undertaken as part of the EURODESAL project on nuclear desalination, currently being carried out by a consortium of four European, and one Canadian, industrials and two leading EU R&D organisations.Major achievements of the project, as discussed in this paper are:
- • Coherent demonstration of the technical feasibility of nuclear desalination through the elaboration of coupling schemes for optimum cogeneration of electricity and water and by exploring the unique capabilities of the innovative nuclear reactors and desalination technologies.
- • Verification that the integrated system design does not adversely affect nuclear reactor safety.
- • Development of codes and methods for an objective economic assessment of the competitiveness and sustainability of proposed options through comparison, in European conditions, with fossil energy based systems.
9.
The seismic qualification of equipment in operating nuclear plants has been identified as a potential safety concern in U.S. Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment in Operating Nuclear Power Plants”. In response to this concern, the Seismic Qualification Utility Group (SQUG), with support from the Electric Power Research Institute (EPRI), has undertaken a program to demonstrate the seismic adequacy of essential equipment by the use of actual experience with such equipment in plants which have undergone significant earthquakes and by the use of available test data for similar equipment. An important part of this program is the development of the methodology and test data for verifying the functionality of electrical relays used in essential circuits needed for plant shutdown during a seismic event. This paper describes the EPRI supported relay testing program to supplement existing relay test data. Many old relays which are used in safe shutdown systems of SQUG plants and for which seismic test data do not exist have been shake-table tested. The testing performed on these relays and the test results for two groups of relays are summarized in this paper. 相似文献
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The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics’ Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units. 相似文献
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In the planned reprocessing plant for spent nuclear fuel elements in Germany components and systems, which form a process-technical unit are integrated in a steel structure called a “module”. In the present paper, the stability and strength of two representative modules under dynamic loading due to earthquake are investigated. Since full scale models of the modules exist, both analytical and experimental investigations are possible. The results of the two investigation methods are in good agreement. Both modules withstand the earthquake loadings with only minor modifications. 相似文献
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The German Basis Safety Concept is an approach which allows the possibility of catastrophic failures to be excluded. It was developed in Germany to render the probabilistic approach unnecessary for safety cases relating to nuclear power plants. The process of evaluation started in 1972, and in 1979 the Basis Safety Concept was officially published and thus became a legal requirement for LWR plants. With appropriate modifications in regard of the particular features of LMFBR, this concept has also been applied to SNR 300. The “Structural Integrity Demonstration Concept” of SNR 300 is based on five principles:
- • - Principle of quality by design and fabrication
- • - Principle of multiple examination
- • - Principle of worst case consideration
- • - Principle of operating surveillance and documentation
- • - Principle of verification and continuous development.
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More and more computers are being used to process and display information to operators who control nuclear power plants. Implementation of computer-generated displays in power plant control rooms represents a considerable design challenge for industry designers. Over the last several years, the Electric Power Research Institute has conducted research aimed at providing industry designers tools to meet this new design challenge. These tools provide guidance in defining more “intelligent” information for plant control and in developing effective displays to communicate this information to the operators. 相似文献
14.
Toshio Tanaka Shigeki Suzuki Yoshitsugu Nekomoto Mamoru Tanaka 《Nuclear Engineering and Design》1994,147(3)
This paper deals with a diagnostic and monitoring system for assessing the integrity of pipe branches, during the operation of the nuclear power plant. This system have been developed under the concept of “easy to use without any sophisticated analysis” and “portable”. The accuracy of the diagnosis is based on the model optimization subsystem, which automatically modifies the numerical vibration model so as to fit its natural frequency to the actual natural frequency. The information obtained by this system may be reflected to a maintenance program of the plant to assure more reliable operation of the plant. 相似文献
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In less than 10 years, the first commercial pressurized water reactor (PWR) plant in Korea will reach its official design life. As part of safety activities, developed countries have already implemented periodic safety review (PSR) or equivalent programs to check and improve the safety of operating nuclear power plants (NPP) during their plant life. At the end of 1999, it was decided by the Korean Atomic Energy Safety Committee to adopt the PSR program and to apply it to Korean operating NPP. Since Kori Unit 1 started the review for the first tentative application of PSR as a model case in May 2000, it is now progressing well. Management of aging is one of the major factors to be considered in PSR and life extension of a nuclear plant. This paper is intended to introduce the regulatory aspect and strategy of Korean PSR. The background and scope of basic PSR guidelines are described, and a summary of technical criteria for aging management, which shows a regulatory direction for PSR, is also presented. 相似文献
16.
R.S. Hart 《Nuclear Engineering and Design》1988,109(1-2)
The CANDU nuclear power system has evolved in a carefully planned and systematic manner over the past 40 years. Thirty-one CANDU power stations are now in commercial operation, or under construction, world wide.Changes in the demands of the world power market, and growing interest in smaller nuclear units led AECL to develop the CANDU 300, which has a net electrical output in the range of 450 MW.AECL initiated design work on the CANDU 300 in 1982. The substantial effort devoted to this program over the past five years has produced a “small” CANDU power station that is economically competitive with both large nuclear and coal fired generating stations.The CANDU 300 makes substantial advances in the areas of station layout, constructability, maintainability, and operability and plant life extension capability while utilizing proven systems, components and concepts. All key components, including steam generators, pressure tubes, fuelling machine, fuel, and coolant pumps are essentially identical to those in service in the very successful CANDU 600 stations.This paper provides an overview of the CANDU 300 with emphasis on factors impacting economics and performance. 相似文献
17.
Analysis of aircraft impact to concrete structures 总被引:1,自引:0,他引:1
Analysis of aircraft impact to nuclear power plant structures is discussed utilizing a simplified model of a “fictitious nuclear building” to perform analyses using LS-DYNA software, representing the loading: (i) by the Riera force history method and (ii) by modeling the crash by impacting a model of a plane similar to Boeing 747-400 to the structure (i.e., “missile–target interaction method”). Points discussed include: (1) comparison of shock loading within the building as obtained from the Riera force history analysis versus from the missile–target interaction analysis, (2) sensitivity of the results on the assumed Riera force loading area, (3) linear versus nonlinear modeling and (4) on failure criteria. 相似文献
18.
Marco Cherubini Nikolaus Muellner Francesco Dauria Gianni Petrangeli 《Nuclear Engineering and Design》2008,238(1):74-80
The University of Pisa was involved in investigations of an Accident Management (AM) procedure based on passive feed water injection. Some experiments were performed to validate this possibility (e.g. in LOBI and Bethsy facilities) and fully analyzed by thermal hydraulic system codes. Recent activities in which the University of Pisa is engaged (also as leader) are focused on VVER-1000 safety analyses. The idea is now to use the acquired knowledge to explore if a procedure based on passive feed water injection is applicable and can provide any benefits to the Russian design pressurised plant.The postulated accident is a station blackout, in such a way only passive systems are available. The proposed AM is based on secondary and primary side depressurisation in sequence. The secondary side depressurisation performed by the BRU-A valves has the scope to feed passively the SGs with the water left in the feed water lines and in the deaerators. The primary side depressurisation, via the PORV, is foreseen to keep the plant at the lowest pressure (to reduce the energy of the system) and to maximize the “grace time” of the plant. Three cases are here considered: no operator action, application of the optimized AM sequence, application of the AM procedure at the last time when it is effective.The intention of this paper is to show that in case of an unlikely event such a SBO the implementation of a strategy based on systems not designed for specific safety application can have a large impact on the “grace time” of the plant. 相似文献
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A review of tests on earthquake-resistant reinforced concrete structural walls is presented. Laboratory tests of isolated walls and construction joints are discussed. Where appropriate, design recommendations are given. The review indicates only few experimental data are available for short walls which are directly applicable to nuclear power plant design. In particular, tests of short rectangular walls subjected to load reversals are needed. Tests are also needed to determine the damping and frequency characteristics of cracked short walls. Analytical and experimental results should be correlated so that the hysteretic response observed in tests can be realistically related to the analytical response “demand” of nuclear power plant structures. 相似文献