日本福岛第一核电站事故源项及后果评价

根据已有的日本福岛第一核电站相关资料,利用美国核管理委员会《轻水堆核电厂事故源项》中的假设条件,计算出事故后安全壳内的放射性源项,综合考虑各种不确定性因素,得出较为保守的环境释放源项。采用美国核管理委员会RG 1.4中大气扩散模式的假设计算大气弥散因子,并应用ICRP 71号出版物F、GR 12号报告等资料中的剂量计算...     

美国核电厂安全辅助目标研究

本文回顾了美国定量安全目标和辅助目标的发展历程,参考NUREG-1150中概率安全分析（PSA）方法,给出了堆芯损坏频率和早期大量放射性释放频率的数学表达式。重点论证了辅助目标和定量安全目标的关系,比较了本文论证方法与NUREG-1860论证方法的不同。研究了美国对新建电站安全目标的要求和PSA技术的发展方向,探讨了其对我国核电厂安全目标和PSA技术发展的借鉴。     

Quantile estimate of the uncertainties of probabilistic safety analysis for objects of the nuclear power industry

Methods for performing comparative analysis of nuclear power plant safety and estimating the residual risk, which are based on analysis of the 95% quantiles of the resulting distributions of the probability density of events which are important for safety, are formulated using an approximate calibration method of quantile estimates of the uncertainties and for the example of the results of a probability analysis of the safety of nuclear power plants in the USA which are presented in the NUREG-1150 report. The basic assumptions of the methods which make it possible to estimate the stationary risk are presented. __________ Translated from Atomnaya énergiya, Vol. 101, No. 3, pp. 167–176, September, 2006.     

国外核燃料工厂核临界事故的经验教训

本文通过对美、英、俄罗斯（前苏联）及日本的核燃料工厂已发生并公开报道的22起核临界事故及美国的一起未遂核临界事故的原因分析，从对核临界事故的认识，核燃料工厂的设计、运行管理和事故的紧急处理以及管理部门的审管方面，分析了我们应该吸取的主要经验教训。     

The NUREG-1150 probabilistic risk assessment for the Grand Gulf Nuclear Station

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Grand Gulf Nuclear Station performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, consequence analyses, and risk results obtained when the results of these analyses are combined with the accident frequency analysis. The offsite risk from internal initiating events was found to be quite low, both with respect to the safety goals and to the other plants analyzed in NUREG-1150. The offsite risk is dominated by short-term station blackout plant damage states. The long-term station blackout group and the anticipated transients without scram (ATWS) group contribute considerably less to risk. Transients in which the power conversion system is unavailable are very minor contributors to risk. The low values for risk can be attributed to low core damage frequency, good emergency response, and plant features that reduce the potential source term.     

The OBE Elimination in Piping Design

The Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) have been considered in the design of nuclear power facilities as required by Appendix A to 10 CFR Part 100. However, it is believed that the elimination of the OBE from the design of nuclear facilities would be necessary for plant optimization since the OBE criterion is too rigid and has excessive conservatism. Studies indicate that alternative piping designs can exhibit reliability and safety levels equal to or greater than the current analysis methods. The alternative rules for the Earthquake Engineering Criteria have been issued by the Appendix S to 10 CFR 50. In the System 80⁺ Design, the USNRC reviewed the alternate analysis methods which were proposed to eliminate the OBE based on the EPRI-URD and concluded that those were acceptable as stated in the NUREG-1462. In the Korean Next Generation Reactor (KNGR) developed as an ALWR, a typical piping model was selected to include ASME Classes 1, 2 and 3 piping and was analyzed according to the current method as well as the alternate analysis method, specified in NUREG-1462, for comparison.     

Incorporation of passive components aging into PRAs

The probabilistic risk assessments being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing a method for selecting risk-significant passive components and including them in probabilistic risk assessments. We demonstrated the method by selecting a weld in the auxiliary feedwater system. The selection of this component was based on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then used the PRAISE computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The calculation included the effects of mechanical loads and thermal transients considered in the design and the effects of thermal cycling caused by a leaking check valve. We modified an existing probabilistic risk assessment (NUREG-1150 plant) to include the possible failure of the auxiliary feedwater weld, and then we used the weld failure probability as input to the modified probabilistic risk assessment to calculate the change in plant risk with time. The results showed that if the failure probability of the selected weld is high, the effect on plant risk is significant. However, this particular calculation showed a very low weld failure probability and no change in plant risk for the 48 years of service analyzed. The success of this demonstration shows that this method could be applied to nuclear power plants.     

The fourth generation of nuclear power

The outlook for nuclear power in the U.S. is currently very bright. The economics, operations and safety performance of U.S. nuclear power plants is excellent. In addition, both the safety and economic regulation of nuclear power are being changed to produce better economic parameters for future nuclear plant operations and the licenses for plant operations are being extended to 60 years. There is further a growing awareness of the value of clean, emissions-free nuclear power. These parameters combine to form a firm foundation for continued successful U.S. nuclear plant operations, and even the potential for new plant construction.     

Development of thermodynamic databases in the system U–Zr–Ce–Cs–Fe–B–C–I–O–H for application to simulating phase equilibria in severe nuclear accidents

In case of severe nuclear accidents involving melt down of nuclear fuels at high temperatures, it is of considerable importance to accurately evaluate the highly-volatizing behavior of fission products (FPs) over multicomponent debris. Particularly, cesium (Cs)- and iodine (I)- bearing chemical species are regarded as notable FPs. In the present work, the authors have generated original thermodynamic databases for the system U–Zr–Ce–Cs–Fe–B–C–I–O–H featuring Cs- as well as I-bearing subsystems, which are contained in oxide, iodide, and metal (including borides and carbides) sub-databases. It has been confirmed that the phase diagrams calculated by the present set of the databases reproduce the corresponding literature data well in various kinds of subsystems of the above multicomponent system. The present set of databases has subsequently been applied to simulate phase equilibria and volatizing behavior of Cs- and I-including species, respectively, in multicomponent debris under specific temperature and atmospheric conditions corresponding to severe nuclear accidents.     

Pneumatic pressure tests of steel containment models — Recent developments

This paper describes the results of recent pneumatic pressure tests of steel containment models. These tests are part of the Containment Integrity Program whose objective is the qualification of methods for predicting containment response during severe accidents and extreme environments. Sandia National Laboratories is conducting this combined experimental and analytical program for the U.S. Nuclear Regulatory Commission (NRC). The long-range plans for the program include the following three containment loading conditions: static internal pressurization, dynamic internal pressurization, and seismic loadings. Steel, reinforced concrete, and prestressed concrete containment types are being considered.In the present experimental effort, models of steel containment structures are being subjected to static internal pressurization. The first set of models are about the size of hybrid-steel containments. Tests of these models are nearly finished. Testing of a large steel model, about of full size, will complete the static pressure experiments with steel models. Analysis of the models is paralleling the experimental effort.The Containment Integrity Program is being coordinated with other NRC programs on potential leakage of penetrations in containments. The results from all of the programs should provide a basis for predicting the structural and leakage behavior of containments during temperature and internal pressure loadings.     

小型堆应急计划区划分方法的探讨

本文总结分析了应急计划区划分中应用的NUREG-0396推荐的方法、概率准则法和风险指引法在小型堆应急计划区划分中的适用性,推荐风险指引法作为较合理的小型堆应急计划区划分方法。讨论了在实际应用中需要关注的应急计划区划分和反应堆设计的相互作用、合理的事故假设和公众心理因素等问题。     

ALWR utility requirements document containment performance requirements

U.S. utilities, with substantial support from international utilities, are leading the industry-wide Advanced Light Water Reactor (ALWR) Program. This program is establishing a technical foundation for the next generation of LWRs through development of a comprehensive set of design requirements for the ALWR in the form of a Utility Requirements Document (URD).The approach in the URD for severe accidents involves two main efforts: (1) accident prevention through intrinsic design characteristics, backed up by reliable safety systems to prevent core damage, and (2) incorporation of design features to ensure severe accident mitigation and containment for the full spectrum of postulated accidents, including core melt accidents.For containment performance, twenty-three severe accident containment challenges are identified, a matrix of design features and operating characteristics is specified to address the challenges, and a preliminary evaluation of the URD indicates that requirements are adequate for addressing each of the challenges. Further, the URD requires evaluation of the containment response to severe accidents.The key conclusions from this effort are: that severe accident challenges are being systematically and explicitly addressed in the design of ALWRs; that margin exists between the loads predicted to result from severe accidents and the Service Level C limits; and that, for a realistic new design basis source term that is expected to bound that from any credible severe accident sequence, the site boundary dose is less than 0.5 rem given the predicted intact containment performance.     

Containment venting as an accident management strategy for BWRS with Mark I containments

A quantitative evaluation of primary containment venting was performed to assess its risk reduction potential. A boiling water reactor with a Mark I containment was evaluated by developing simplified containment event trees for its risk dominant sequences. Risk results were benchmarked with those from the NUREG-1150 risk rebaselining effort, and sensitivity studies then were performed. It was found that for station blackout sequences, containment venting by itself does not significantly reduce overall risk. For sequences involving loss of long-term decay heat removal or failure to scram, however, venting is potentially an important mechanism in preventing or delaying core melting. Subsequent studies show that when venting is combined with other potential containment improvements, there is a large potential for risk reduction.     

Spent fuel transportation risk assessment: conclusions

Abstract

The US Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. The study reached the following findings. First, the collective dose risks from routine transportation are vanishingly small. These doses are about four to five orders of magnitude less than collective background radiation doses. Second, the routes selected for this study adequately represent the routes for spent nuclear fuel transport, and there was relatively little variation in the risks per kilometre over these routes. Third, radioactive material would not be released in an accident if the fuel is contained in an inner welded canister inside the cask. Fourth, only rail casks without inner welded canisters would release radioactive material, and only then in exceptionally severe accidents. Fifth, if there were an accident during a spent fuel shipment, there is less than one in a billion chance the accident would result in a release of radioactive material. Sixth, if there were a release of radioactive material in a spent fuel shipment accident, the dose to the maximally exposed individual would be <2 Sv (200 rem) and would not cause an acute fatality. Seventh, the collective dose risks for the two types of extraregulatory accidents (accidents involving a release of radioactive material and loss of lead shielding) are negligible compared to the risk from a no release, no loss of shielding accident. Eight, the risk of loss of shielding from a fire is negligible. Ninth, none of the fire accidents investigated in this study resulted in a release of radioactive material. Based on these findings, this study reconfirms that radiological impacts from spent fuel transportation conducted in compliance with NRC regulations are low. In fact, this study’s radiological impact estimates are generally less than the already low estimates reported in earlier studies. Accordingly, with respect to spent fuel transportation, this study reconfirms the previous NRC conclusion that the regulations for transportation of radioactive material are adequate to protect the public against unreasonable risk.     

Comparative population dose risks from nuclear fuel cycle closure and renewal of the commercial nuclear energy alternative in the U.S.

The debate over a large expansion of commercial nuclear energy for electricity production in the U.S., termed a “nuclear renaissance,” has most recently focused on the issues of spent nuclear fuel transportation and the closing of the once-through nuclear fuel cycle through the licensing, construction, and operation of the national spent nuclear fuel repository at Yucca Mountain, Nevada. While such a commercial nuclear energy expansion is postulated to have environmental, climate, resource utilization, and economic benefits, the fundamental issue for typical U.S. citizens about nuclear energy concerns the potential for exposure to ionizing radiation. Two generations of U.S. citizens have experienced public and media “education” that has heightened their primal fears of ionizing radiation from commercial nuclear energy. In such an environment, comparing the risks of radiation doses from commercial nuclear energy fuel cycle closure and further nuclear energy expansion with ionizing radiation population doses experienced year after year, decade after decade from non-nuclear (conventional) industries seems worthwhile for use in achieving stakeholder education and concurrence. The U.S. National Academy of Sciences (NAS) has recently performed its own landmark risk assessment of spent fuel transport in the U.S., demonstrating the guiding principles and methods for use in comparative risk assessments involving radiation dose considerations. Using the NAS assessment approach, this paper broadens its application to the full consideration of the risk of nuclear fuel cycle closure and renewal of the commercial nuclear energy alternative in the U.S., to evaluate the ionizing radiation dose risks of such expansion compared to those routinely accepted for non-nuclear industries by policy makers and the public. The 50-year collective dose risk from the total commercial nuclear fuel cycle, even if the U.S. triples its installed nuclear capacity, transports spent fuel to Yucca Mountain, and operates the Yucca Mountain repository as planned, is shown to be in the range of 3.1-million person-cSv; for five selected non-nuclear industries, the corresponding 50-year collective dose risk exceeds 1 billion person-cSv, a more than 300 times greater risk. A key step towards renewing the commercial nuclear energy alternative, then, is to use this knowledge for education of various stakeholder parties.     

Development of a consensus standard for verification and validation of nuclear system thermal-fluids software

With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal–hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the jurisdiction of the V&V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. Although software verification will be an important and necessary part of the standard, much of the initial effort of the committee will be focused on the validation of existing software and new models that could be used in the licensing process. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general requirements for the proposed V&V 30 Standard, which includes: (a) applicable NRC and other regulatory requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas-Cooled Reactors, it is anticipated that the practices and procedures developed for this Standard can eventually be extended to other nuclear and non-nuclear applications.     

Approach to accident management in RBMK-1500

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK.Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed.This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account.     

Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles

As a result of feedwater nozzle cracking observed in Boiling Water Reactor (BWR) plants, several design modifications were implemented to eliminate the thermal cycling that led to crack initiation. BWR plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of this, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0619, which established periodic ultrasonic testing (UT) and liquid penetration testing (PT) requirements. While these inspections are useful in confirming structural integrity, they are time consuming and can lead to significant radiation exposure to plant personnel. In particular, the PT requirement poses problems since it is difficult to perform the inspections with the feedwater sparger in place and also leads to additional personnel exposure. Clearly, an inspection and monitoring program that eliminates the PT examination and still verifies the absence of surface cracking would be extremely valuable in limiting costs as well as radiation exposure. This paper describes a program involving the application of advanced UT techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles. The inspection methods include: (1) scanning with optimized transducers and techniques from the outside vessel wall surface to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that show crack location, size, and maximum depth of penetration into the nozzle inner surface. These techniques have been developed to the point where they are now considered a reliable alternative to the liquid penetrant requirements of NUREG-0619. An important supplement to the UT program is the use of automated fatigue, leakage and crack growth monitoring to verify the absence of cracking. This approach provides for a continuous assessment of the integrity of the nozzle structure by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage and performing crack growth predictions based on actual thermal duty. Collectively, the monitoring and inspection program provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.