美国三级PSA发展过程及研究进展

公众风险是验证核电厂安全性能的重要指标。美国1957年发布的WASH-740中开始引入反应堆风险分析的概念;1972年,WASH-1400第一次系统性地完成了三级PSA,并开发了CRAC系列分析程序;1990年,NRC发布NUREG-1150,标志着PSA应用进入成熟期;1995年,正式在法规层面上确定了PSA地位;始于2007年的SOARCA研究计划在SARP研究基础上,结合最新研究进展利用改进的三级PSA工具对两个核电厂重新进行了评估;2011年,考虑最新的技术进步和福岛事故影响,计划重新启动完整三级PSA研究,以考虑如乏燃料水池工况、群堆工况、外部事件等新的认识。     

中系技术规格书转化的若干问题研究

在大亚湾核电站等M310型核电机组参照美国《西屋核电厂标准技术规格书》（NUREG-1431）第4版将技术规格书由法系转为中系的过程中发现，NUREG-1431第4版管理体系中缺少对多个安全功能或安全相关系统和设备同时不可用的管理。为此，本文充分研究我国M310型核电机组多个安全功能或安全相关系统和设备同时不可用的管理现状，对比技术规格书由法系转为中系带来的风险影响，总结管理方式变化产生的安全问题，最后提出了针对性的工作建议。本文研究对国内M310型核电机组技术规格书由法系转为中系相关工作具有借鉴意义。     

溶液堆内部事件概率安全分析框架研究

液体燃料反应堆（简称溶液堆）与传统固体燃料反应堆在安全设计和运行特性等方面存在重大差异,无法仅按照现有以确定论为核心的设计方法进行安全设计,必须在设计之初引入概率安全分析（PSA）技术。由于燃料形态、安全屏障及缓解系统等与固体燃料反应堆的差异,传统以堆芯损坏为核心的反应堆PSA技术无法直接适用于溶液堆。在调研国内外传统研究堆、溶液堆及乏燃料后处理厂相关要求及分析技术后,以我国正在研发的医用同位素试验堆为对象,提出了溶液堆PSA安全目标,并建立了PSA技术框架,为该类型反应堆PSA的开展和安全审查奠定基础。     

概率安全评价在核能安全分析领域的应用和发展

概率安全评价（PSA）是核能安全分析领域的两大分析方法之一。本文从PSA概念入手,首先从理论基础、分析视角等多个方面比较了确定论和概率论2种分析方法的差异;其次,梳理PSA在核能安全分析领域的历史进程,通过回顾PSA在技术和法规上的变化,展示了PSA与核能安全在提升过程中相互促进的关系;再次,阐释PSA技术在风险量化预测、平衡安全设计、安全决策、安全监管方面的应用,并通过华龙一号（HPR1000）的实例展示了PSA在核能安全分析中的具体应用方式。最后,对PSA技术未来的发展方向进行了预测,指出确定论和概率论2种分析方法将深入融合,PSA分析从安全目标向任务目标转移、从静态向动态转换、从认知向感知转换的发展方向。      

基于PSA技术的核电厂数字化仪控系统可靠性设计及应用

传统意义上核电厂数字化仪控系统主要依靠提升设备的可靠性来满足电厂安全目标。随着监管要求的逐步提高,在提升设备可靠性基础上,基于概率论技术的设计手段逐步成为核电厂安全设计新的研究方向。本文应用概率安全评价（PSA）技术,对典型电厂始发事件进行分析及研究,之后对仪控设计方案整体进行PSA建模,再将其置于电厂PSA模型中,通过定量评估分析,识别薄弱环节,给出优化改进措施。在此基础上提出了一套确保核电厂仪控系统满足整体安全目标的可靠性设计流程。      

西安脉冲堆定量安全目标初步研究

为了提高西安脉冲堆运行安全管理水平,结合西安脉冲堆纵深防御特点及现有运行安全水平,以概率安全分析结果为依据,分析提出了西安脉冲堆安全目标体系构成和安全目标定量化数值。研究表明,西安脉冲堆安全目标体系应补充概率安全目标和辐射防护定量约束目标,概率安全目标包含堆芯损伤频率值、少量放射性释放频率值。提出的定量化安全目标建议为:堆芯损伤频率限值1×10~(-5)/堆年,少量放射性释放频率限值1×10-~(7)/堆年;正常工况下,公众有效剂量不超过0.1 m Sv·a~(-1),工作人员有效剂量不超过2 m Sv·a~(-1);事故工况下的定量辐射防护目标还需要进一步的研究。     

基于概率安全评价的技术规格书优化

概率安全评价(PSA)作为一个风险分析的方法,在国内外核电厂风险管控中得到了越来越广泛的应用,美国部分核电厂已经开发并实施了基于PSA分析的技术规格书。但对于重水堆的技术规格书优化,在国际上目前尚无该方面的经验,本文参考美国相关的技术导则与实践,以秦山核电厂三期2号停堆系统试验频率为例进行优化分析,评价该方法在重水堆机组的适用性,探索重水堆技术规格书优化的方法。     

连载讲座 轻水堆概率安全评价（PSA）入门 第3讲 内部事件1级PSA

1 前言 本文为概率安全评价（PSA）第3讲，主要讨论运行核电站内部初因事件所涉及的1级PSA。正如第1讲阐述的，1级PSA用于研究未造成堆芯损坏的事故工况，并评价其发生频率。根据1级PSA的评价结论和堆芯损坏频率可弄清楚重要的事故状态、设备故障和人员差错等的影响。另外，如第2讲所示，1级PSA技术被应用于各种安全管理、安全规章制度的领域。以下对1级PSA的方法进行叙述，关于各种方法的详细说明、实施例以及停堆工况的PSA，请参阅本文所附的参考文献。     

船舶核动力装置安全目标的初步研究

安全目标是核电厂进行安全评价的重要基础和判定准则,对安全评价工作具有非常重要的影响.目前,核电厂安全目标的认识和制定已经经历了较长的时间,形成了以国际原子能机构(IAEA)和美国核管会(NRC)为主的两大体系.文章概要地介绍了两个组织所确定的定性和定量安全目标,以及我国核电厂安全目标的发展和应用现状.最后,在吸收上述经...     

核电厂起火频率分析

通过研究美国核管理委员会(NRC)和电力研究院(EPRI)的《核电厂的火灾概率风险评价方法》,介绍了核电厂内部火灾概率安全评价(PSA)过程中,各类点火源起火频率的分析方法和步骤。以大亚湾核电厂的主变压器为例,介绍了起火频率的具体分析过程。经定量计算分析,大亚湾核电厂主变压器的起火频率是4.32×10-3/(堆.年),是反应堆堆芯损坏频率(CDF)的203倍。一旦起火,发生破坏性火灾的概率高达83%。     

关于核电厂安全目标的确定问题

介绍了有关核电厂定性安全目标、定量安全目标和概率安全目标的定义,讨论了确定这些目标时所涉及的一些问题.     

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.     

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.     

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.     

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.     

Current perspectives on nuclear power plant risks and the use of risk-based information in regulation

For the past five years, the U.S. Nuclear Regulatory Commission has supported extensive studies of severe accidents. One outcome of this work is a set of advanced method for analyzing the probabilities, source terms, consequences, and risks of such accidents. These methods are being applied to a set of six U.S. commercial nuclear power plants, covering a wide spread of nuclear steam supply systems and containment designs. This work is to be documented in the Reactor Risk Reference Document, NUREG-1150, and supporting contractors reports. The methods being used for NUREG-1150, and some initial plant results, are briefly described in this paper.     

核安全目标制定与评估中的若干问题

介绍美国核管会（NRc）关于核电站运行的安全目标及其发展过程，讨论在安全目标的制定和评估中所涉及的主要问题，包括事故预防与事故缓解的平衡、不确定性的处理、安全目标的实施等，同时概要介绍了其他一些国家和机构所提出的安全目标，最后对安全目标的发展和实际应用前景进行了讨论。     

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。     

华龙一号核电厂公众辐射剂量优化设计目标值研究

基于我国新建三代压水堆核电机组华龙一号运行状态下的流出物设计与现实排放源项,结合机型的排放特点、沿海与内陆厂址的不同环境和气象等条件,对新建核电厂公众剂量优化设计目标值进行了研究。研究结果表明,在我国华龙一号的设计方案下,对于滨海厂址条件所能够达到的公众剂量优化设计目标值可以达到与欧美国家体系相当的水平。但由于厂址环境条件的差异,对内陆核电厂还需要结合工艺系统的改进开展进一步深入研究工作。对流出物中氚和C-14的排放还需要进一步开展经验反馈积累,在此基础上给出更符合未来运行状态的辐射剂量评价结果。本研究还对华龙一号后续的排放源项和评价方法等方面给出了建议,对进一步加强华龙一号的环境友好性和先进性具有积极作用。