百万千瓦级压水堆核电厂停堆PSA研究

对百万千瓦级核电厂的停堆运行事故风险进行内部事件1级概率安全评价(PSA),并根据不同的停堆进程分别建立停堆PSA模型,分析经历LOI-RRA水位对电厂风险水平构成的影响。分析结果表明停堆工况下的电厂风险不可忽视,在冷停堆工况下经历LOI-RRA水位导致堆芯损坏频率明显增加。     

压水堆核电站停堆工况硼失控稀释事故风险研究

以我国大亚湾核电站为例,对压水堆电站停堆工况下硼失控稀释的潜在事故谱进行了系统的分析并归类,然后采用PSA方法并基于法国核电站750堆年运行经验反馈数据,对其潜在事故风险进行了定性和定量评价,提出了有针对性的降低事故风险的建议和措施。     

压水堆核电站停堆工况硼失控稀释事故风险研究

以我国大亚湾核电站为例,对压水堆电站停堆工况下硼失控稀释的潜在事故谱进行了系统的分析并归类,然后采用 PSA 方法并基于法国核电站 750 堆年运行经验反馈数据,对其潜在事故风险进行了定性和定量评价,提出了有针对性的降低事故风险的建议和措施。     

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。     

低功率及停堆工况PSA中的电厂运行状态分析

电厂运行状态(POS)分析的目的是将核电厂低功率停堆运行这一连续的动态过程离散化,这是用事件树表示发展事故序列的必要条件。以某300 MW参考核电厂的设计、运行经验、操作规程等基础做为参考,采用相关准则进行详细的POS分析,得到合理的POS,并根据该参考电厂实际运行情况计算得到每个POS的持续时间。这项工作为开展低功率及停堆工况PSA奠定了重要的基础,其分析方法和內容为国內开展此项工作提供了参考。     

EPR的二级概率安全评价

EPR设计广泛采用了概率安全分析（PSA）作为确定论分析的补充。PSA采用三级分析评价电厂运行所带来的风险。1级PSA用于导致堆芯损坏熔化事件的风险评价，并确定对风险有贡献的事件、系统失效及运行错误。2级PSA用于评价裂变产物从电厂释放到环境的风险，并对严重事故导致的放射性释放（通常称为源项）的频率和大小进行量化分析。3级PSA对事故所导致的放射性释放对社会造成的危害进行量化分析，也就是对健康和对食物链污染的可能影响。     

秦山二期扩建工程低温超压事故PSA分析

低温超压事故在电厂停堆期间发生频率较高,并有可能导致堆芯熔化,是停堆工况下一个重要的安全问题。本文对一回路发生低温超压事故进程进行研究和分析,参考相关资料建立事件树,进行定量化计算,得到低温超压事故导致的堆芯损坏频率,并进行简单的结果评价。     

西安脉冲堆满功率运行工况内部始发事件一级概率安全评价

针对西安脉冲堆(XAPR)2 MW满功率运行工况,建立了内部始发事件一级概率安全评价(PSA)模型,对始发事件识别、事故序列分析及可靠性数据处理等进行了研究。应用小事件树-大故障树方法,在Risk Spectrum平台上完成XAPR堆芯损伤事故序列的定量分析。结果表明,XAPR内部事件导致的堆芯损伤频率(CDF)为4.14×10~(-6)/(堆·年),对CDF贡献最大的为堆水池堆芯高度处大破口失水事故,支配性事故序列是大破口失水事故后紧急排水系统失效。研究结果证明XAPR具有较高的安全性。     

49-2游泳池式反应堆超设计基准事故的筛选与分析

为保证49-2游泳池式反应堆在超寿期下的安全运行,需进行超设计基准事故分析。由于难以采用概率安全评价(PSA)方法进行分析,所以本文无条件假设最严重事故来得到一保守结果。主要分析了全厂断电下未能紧急停堆的预期瞬变(ATWS)、水平孔道断裂和停堆后堆芯完全裸露的事故,以及应急能力。结果表明:在全厂断电ATWS下堆芯是安全的;水平孔道断裂及其他因素造成失水时,只要2.5h内堆芯不裸露即可保证燃料元件不熔化;非能动破坏虹吸能力和多样的应急补水方式能保证堆芯不裸露。     

火灾情景下的人员可靠性分析

国内外各核电厂火灾概率安全评价（PSA）表明,人员操作对火灾情景下的电厂风险有重要影响,因此,有必要采用系统的人员可靠性分析（HRA）方法来评价火灾情景下的人员失误概率。本文阐述了HCR/ORE和CBDTM模型的基本理论和在火灾情景下的特殊考虑。将HCR/ORE和CBDTM方法与THERP方法相结合应用于火灾情景下的人员可靠性分析,并进行了实例分析。为建立更符合工程实际的火灾PSA模型奠定了基础。     

大亚湾核电站停堆工况风险研究

基于传统ＰＳＡ方法学（适用于功率运行工况）及核电站停堆工况特征,提出了一套停堆ＰＳＡ特征方法,包括电站运行状态离散法,分阶段评价和主逻辑故障树评价。将该方法应用于大亚湾核电站（ＧＮＰＰ）停堆工况ＰＳＡ研究,得到了较真实反映ＧＮＰＰ实际情况的结果。研究结果对ＧＮＰＰ的停堆运行和管理有实际应用价值,以我国今后核电站设计、运行及管理也有现实意义。     

Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during cooldown to cold shutdown, and in the validity of a two-tier calculational method. The results have been directly used in updating the plant shutdown PSA, by changing the success criteria for core cooling during cooldown of the plant and showing a reduction in overall risk.     

核电站停堆工况 PSA 研究

停堆工况下核电站概率安全评价技术及应用是ＰＳＡ研究热点之一，已受到研究单位、核安全管理当局和业主的普遍关注。文章简要介绍了国外有关停堆ＰＳＡ的进展情况，建议在我国开展停堆ＰＳＡ研究。     

Risk-informed decision-making related to the on-line maintenance

On-line maintenance (OLM) represents the term, which includes testing and maintenance that is performed when the main generator of the nuclear power plant is connected to the grid. OLM on one side helps to decrease the number of activities, which would be performed during the scheduled outage, but on the other side it may contribute to a different level of risk, if the activity is performed when the plant is operating. If the risk of OLM during the power operation is much larger than the risk of similar activity performed during the shutdown the OLM may not be the desired strategy. Additionally, if the risk of OLM during the power operation is not larger than the risk of similar activity performed during the shutdown, the OLM would be the preferred strategy.The objective of the paper is to show risk evaluations of selected OLM activities of selected real plants and to examine if the current practices are suitable for the new nuclear power plants.The results of risk evaluations of OLM and their comparison show that the criteria or guidelines developed for existing plants are not completely suitable for new plants. The new power plants with expected lower risk measures, which can be lower for more than couple of orders of magnitude compared to existing plants, would be able to deal with OLM in plant configurations which would increase the risk for orders of magnitude, but would still be acceptable in terms of risk, if the existing criteria or guidelines are used. Results suggest that the risk guidelines for the OLM should be updated for their use in the case of new nuclear power plants.     

EPR机组反应堆厂房空气污染评估

为评估EPR机组功率运行和停堆期间反应堆厂房的空气污染水平,本文介绍了空气污染评估方法,并建立了H-3的空气浓度估算模型;估算了功率运行期间和停堆期间反应堆厂房设备间和工作间的空气污染水平。评估结果显示,在停堆期间,反应堆厂房空气污染的主要核素是H-3,大约为0.07 DAC,其导致的大修集体剂量可以达到56人·mSv。     

Computation and measurement of calandria tube sag in pressurized heavy water reactor

Calandria tubes and liquid injection shutdown system nozzles in a pressurized heavy water reactor are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with liquid injection shutdown system tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measurement probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sag between both tubes in the reactor. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted.     

Toward risk-based control of nuclear power plant configurations

This paper presents an analysis of risks associated with component outage configurations during power operation of a nuclear power plant and discusses approaches and strategies for developing a risk-based configuration control system. A configuration, as used here, is a set of component states. The objective of risk-based configuration control is to detect and control plant configurations using a risk perspective.The configuration contributions to core-melt frequency and core-melt probability are studied for two plants. Some equipment configurations can cause large core-melt frequency increases and there are a number of such configurations that are not currently controlled by technical specifications. However, the expected frequency of occurrences of the impacting configurations is small and the actual core-melt probability contributions are also generally small. Effective strategies and criteria for controlling configuration risks are presented. Such control strategies take into consideration the risks associated with configurations, the nature and characteristics of the configuration risks, and also the practical considerations such as adequate repair times and/or options to transfer to low risk configurations. Alternative types of criteria are discussed that are not overly restrictive to result in unnecessary plant shutdown, but rather motivate effective test and maintenance practices that control risk-significant configurations to allow continued operation with an adequate margin to meet challenges to safety.     

Risk-Based,genetic algorithm approach to optimize outage maintenance schedule

A huge number of components are typically scheduled for maintenance when a nuclear power plant is shut down for its planned outage. Among these components, a number of them are risk significant so that their operability as well as reliability is of prime concern. Lack of proper maintenance for such components during the outage would impose substantial risk on the nuclear power plant (NPP) operation.