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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。 相似文献
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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。 相似文献
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运用故障树分析方法,对广东大亚湾核电站(GNPP)厂用电力系统的可靠性作了分析。建造了电力系统6.6kV交流应急母线(LHA)、220V交流不间断电源母线(LNE)和125V直流电源母线(LBA)的失电故障树。利用SETS程序及法国标准900MW压水堆核电站200堆·年运行经验反馈的可靠性数据,对电力系统的可靠性作了定性、定量分析。给出了电力系统故障树支配性最小割集和顶事件的发生概率,并对支配性最小割集作了描述和分析。 相似文献
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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。 相似文献
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对百万千瓦级核电厂停堆运行事故进行内部事件1级概率安全评价(PSA),根据不同的停堆进程分别建立停堆PSA模型,分析经历余热排出系统(RRA)低运行区间(LOI-RRA)水位对电厂风险水平构成的影响;同时采用事故系列先兆标准电厂风险分析模型人员可靠性分析(SPAR-H)方法进行人员可靠性分析,评价其定量化结果的适用性。分析结果表明,停堆工况下的电厂风险不可忽视,在停堆工况下的事故规程有待完善之处,冷停堆工况下由LOI-RRA水位导致堆芯损坏频率明显增加,人因失误是造成停堆高风险的关键因素。 相似文献
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CANUDU重水堆燃料管理 总被引:1,自引:1,他引:0
论述秦山三期核电站所采用的CANDU-6反应堆的燃料管理,CANDU堆的换料是带功率刊物 ,这一特征使得它的堆内燃料管理与必须停换料的反应堆有明显的不同。CANDU堆燃料管理有设计和运行两方面的内容。 相似文献
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核电站反应堆循环停堆日期预测及循环长度的评价都是为燃料管理提供设计输入.本文介绍了两种循环停堆日期预测方法,并指出了其适用范围;同时介绍了循环长度的标定方法,并用该方法评价了几个循环的理论循环长度,最后分析了标定误差. 相似文献
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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。 相似文献
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Life extension is investigated as a safeguard assessment for the stability on the operation of the nuclear power plants (NPPs). The Cobb-Douglas function, one of the production functions, is modified for the nuclear safeguard in NPPs, which was developed for the life quality of the social and natural objects. Nuclear Safeguard Estimator Function (NSEF) is developed for the application in NPPs. The cases of NPPs are compared with each other in the aspect of the secure performance. The results are obtained by the standard productivity comparisons with the designed power operations. The range of secure life extension is between 1.008 and 5.353 in 2000 MWe and the range is between 0.302 and 0.994 in 600 MWe. So, the successfulness of the power operation increases about 5 times higher than that of the interested power in this study, which means that the safeguard assessment has been performed in the life extension of the NPPs. The technology assessment (TA) is suggested for the safe operation which is an advanced method comparing conventional probabilistic safety assessment (PSA). 相似文献
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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. 相似文献
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核电厂运行经验反馈和概率安全分析表明,核电厂在停堆状态下具有相当大的堆芯熔化的潜在风险。本文叙述了核电厂运行经验反馈,概率安全分析和事故分析的结果,以及相关的措施和研究课题,特别涉及到停堆状态下的非可控硼稀释事故的维修冷停堆下推动余热排出系统。 相似文献
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随着核电厂安全分析方法的不断发展,结合传统确定论分析与概率风险评价(PSA)的风险指引型安全分析方法逐渐引起安审当局和核电业主的广泛关注。本文基于国际上风险指引型分析方法在其他领域的应用现状,提出了风险指引的大破口失水事故(LBLOCA)分析方法,并重新评估了CPR1000核电厂的堆芯燃料包壳峰值温度(PCT)裕量。在PSA分析中,识别并量化了LBLOCA发生后可能发生的162个事件序列,并采用确定论现实分析方法(DRM)对筛选出的18个概率较大的事件序列进行了计算分析。然后通过期望值评估法和特定序列覆盖法对LBLOCA的PCT裕量进行了评估。结果表明,本文方法下LBLOCA的PCT裕量约为36~55 ℃,相比于传统的DRM裕量提升了16~35 ℃。 相似文献