非能动压水堆核电厂乏燃料池风险评价

以非能动压水堆核电厂为研究对象,对可能引起乏燃料损伤的内部事件进行了风险评价。采用PSA软件RiskSpectrum建立事件树和故障树模型,进行乏燃料损伤频率（FDF）定量化。结果表明：在所有工况下总的FDF为2.05×10^-9/(堆•年),远小于堆芯的损伤频率（约2.41×10^-7/(堆•年)）;即使在放射性完全释放的假设下,乏燃料损伤导致的大量放射性释放频率仍较堆芯损伤导致的大量放射性释放频率（约2.38×10^-8/(堆•年)）低1个量级;由于非能动压水堆核电厂有多重预防缓解措施以应对乏燃料池（SFP）事故,SFP风险远低于堆芯风险,可实现核安全导则的安全目标。     

压水堆核电厂蒸汽发生器传热管道破裂事故源项的计算分析

当压水堆核电厂发生事故后,带有放射性的核素会通过破损处释放到环境中,从而危害核电厂周边环境及相关人员的安全,因此对事故后释放到环境中的放射性源项分析,对于核电厂的辐射防护具有重要意义。本文根据事故发生的频率以及后果严重程度,选取蒸汽发生器传热管破裂事故(Steam Generator Tube Rupture,SGTR)进行分析。事故分为事故前碘尖峰释放和事故并发碘尖峰释放两种事故工况,建立事故后放射性核素迁移和扩散计算模型,同时使用先进压水堆AP1000参数进行计算验证,并重点关注惰性气体和挥发性核素碘在环境中的放射性活度。计算结果显示:使用文中计算模型计算的放射性源项与设计源项比较一致,在两种工况下,惰性气体的释放活度与设计源项吻合较好,但碘的释放活度有明显差别。     

压水堆核电站蒸汽发生器传热管破损监测和破裂事故

文中指出,蒸汽发生器排污取样分析及主蒸汽管道外~(16)N 监测是蒸汽发生器传热管破损监测的主要手段。文中还介绍了发生传热管破损后的事故过程及其处理措施。最后对传热管破损事故的审批办法作了介绍.     

EPR缓解SGTR事故的设计特点

蒸汽发生器传热管破裂(SGTR)事故可能造成安全壳旁通,是一个特殊而重要的设计基准事故。本文归纳了EPR缓解SGTR事故的主要设计特点:(1)中压安注(MHSI)泵关闭扬程低于主蒸汽安全阀(MSSV)开启整定值,避免MSSV开启;(2)大气旁排系统(VDA)通过降低其整定值自动启动部分冷却,使一回路快速冷却、降压;(3)蒸汽发生器排污系统(APG)增加转移管线,有助于以排污和蒸汽排放组合方式最终冷却、降压。这些EPR设计特点可供CPR1000核电厂系统设计改进参考。     

船用核动力SGTR事故安全分析

本文论述了船用核动力装置蒸汽发生器传热管断裂事故（SGTR）安全分析的重要性。并以陆奥号核动力商船为例，运用事件树分析技术，对SGTR事故进行了安全分析。得出了事故后可能导致堆芯熔化的事故序列，确定了堆芯熔化数学模型，并进行了定量化分析。最后根据对支配性事故序列和各题头事件重要度的分析，提出了降低SGTR事故导致堆芯熔化发生概率应采取的相应措施。     

压力诱发SGTR的概率计算方法探讨

诱发蒸汽发生器传热管破裂现象对核电厂堆芯损伤和放射性早期大量释放风险有非常大的影响。准确地对诱发蒸汽发生器传热管断裂概率进行计算和定值对正确认知核电厂的核安全风险非常重要。文章调研了已有压力诱发蒸汽发生器传热管破裂概率的取值计算方法,通过对不同取值计算方法的对比分析,结合国内实际情况,提出了一套较为合理可行的压力诱发蒸汽发生器传热管破裂概率的计算方法。文章推荐的诱发蒸汽发生器传热管破裂数据采集和分析计算方法为后续国内核电厂概率安全分析应用和安全监管提供了参考。     

先进压水堆核电厂应急计划区探讨

探讨了影响核电厂应急计划区大小的主要因素并阐述了作者的技术见解,提出了在先进压水堆核电厂应急计划区范围测算过程中,严重事故截断概率取10^(-8)的建议。     

压水堆核电厂事故工况下最大上充流量的数值研究

使用计算流体软件Flowmaster 7.5模拟了某压水堆核电站的化学和容积控制系统(RCV)以及安全注入系统(RIS)。通过将正常工况下的模拟计算结果与设计参考值进行对比,验证了模型的可靠性。使用该模型预测了发生蒸汽发生器传热管破裂事故(SGTR)后安全注入模式转为上充模式时最大上充流量与一回路压力的关系,并定量比较了采用不同数量的低压安注泵为上充泵增压时最大上充流量和相应主泵密封注入流量与一回路压力的关系。     

先进压水堆非能动安全系统研究进展

介绍了我国先进压水堆非能动安全系统研究进展及国内外先进压水堆非能动安全系统研究发展状况,指出我国非能动安全系统研究的发展方向是进行新一代1000MW级压水堆非能动安全系统的研究。     

先进压水堆核电厂安全壳地坑滤网设计

核电厂的地坑滤网堵塞问题是目前世界核能领域比较关注的重要安全问题之一,先进压水堆的设计也充分体现了改善地坑性能的最新研究成果。本文介绍了在我国已经开工建设的先进压水堆核电厂（AP1000以及EPR）的地坑滤网设计,可以为我们解决地坑滤网堵塞问题提供有益的借鉴。     

Success criteria analysis in support of probabilistic risk assessment for nuclear power plants: application on SGTR accident

Success criteria analysis (SCA) bridges the gap between deterministic and probabilistic approaches for risk assessment of complex systems.To develop a risk model,SCA evaluates systems behaviour in response to postulated accidents using deterministic approach to provide required information for the probabilistic model.A systematic framework is proposed in this article for extracting the front line systems success criteria.In this regard,available approaches are critically reviewed and technical challenges are discussed.Application of the proposed methodology is demonstrated on a typical Westinghouse-type nuclear power plant.Steam generator tube rupture is selected as the postulated accident.The methodology is comprehensive and general;therefore,it can be implemented on the other types of plants and complex systems.     

Success criteria analysis in support of probabilistic risk assessment for nuclear power plants: application on SGTR accident

Success criteria analysis (SCA) bridges the gap between deterministic and probabilistic approaches for risk assessment of complex systems.To develop a risk model,SCA evaluates systems behaviour in response to postulated accidents using deterministic approach to provide required information for the probabilistic model.A systematic framework is proposed in this article for extracting the front line systems success criteria.In this regard,available approaches are critically reviewed and technical challenges are discussed.Application of the proposed methodology is demonstrated on a typical Westinghouse-type nuclear power plant.Steam generator tube rupture is selected as the postulated accident.The methodology is comprehensive and general;therefore,it can be implemented on the other types of plants and complex systems.     

Probabilistic safety analyses developments resurgence in Bulgaria—Kozloduy nuclear power plant example

This paper presents some of the main technical features and insights of the Kozloduy nuclear power plant (NPP) units 5 and 6 probabilistic safety analysis (PSA) level 1. Probabilistic analyses and their applications in Bulgaria were given further impetus in recent years. More than 17 years after the first PSA study in Bulgaria in 1992 today probabilistic analyses receive increasing attention and application than ever before. The Bulgarian regulatory body (BNRA) is also interested in expanding their capability of reviewing and using PSA in plant safety assessments. In November 2008 within the framework of the program financed by European Union (PHARE), a project for assisting the BNRA in establishing the regulatory requirements on the base of PSA was completed. One of the objectives of this project was performance of the independent review of Kozloduy NPP units 5 and 6 PSA. This review was a new impulse for the authors to present in more details of Kozloduy NPP probabilistic assessment studies in the present paper.     

Application of thermal comfort theory in probabilistic safety assessment of a nuclear power plant

Human factor errors in probabilistic safety assessment(PSA) of a nuclear power plant(NPP) can be prevented using thermal comfort analysis.In this paper,the THERP+HCR model is modified by using PMV (Predicted Mean Vote) and PPD(Predicted Percentage Dissatisfied) index system,so as to obtain the operator cognitive reliability,and to reflect and analyze human perception,thermal comfort status,and cognitive ability in a specific NPP environment.The mechanism of human factors in the PSA is analyzed by operators of skill,rule and knowledge types.The THERP+HCR model modified by thermal comfort theory can reflect the conditions in actual environment,and optimize reliability analysis of human factors.Improving human thermal comfort for different types of operators reduces adverse factors due to human errors,and provides a safe and optimum decision-making for NPPs.     

AP1000核电建设安全标准化管理

2002年国家颁布实施了《中华人民共和国安全生产法》,出台了一系列的方针和政策,加大了安全生产监管力度。安全标准化是继安全评价、安全生产许可证制度之后,在安全监管方面采取的一项重大举措,也是一项治本之策。安全标准化工作是一项战略性、长期性、基础性的工作,是实现企业安全科学管理,提高企业本质安全的基本途径。海阳AP1000核电,其模块化、平行法施工及开顶法施工颠覆了传统先土建后安装的建造的理念,缩短建设周期,与此同时特大型模块运输、吊装,立体交叉施工显著增加,对安全管理提出了更高的要求。本文结合海阳AP1000核电工程特点及安全管理难点,提出AP1000核电安全标准化管理的思路和措施,对AP1000核电建设安全管理工作有借鉴意义。     

Probabilistic risk assessment in nuclear power plant regulation

A specific program is recommended to utilize more effectively probabilistic risk assessment in nuclear power plant regulation. It is based upon the engineering insights from the Reactor Safety Study (WASH-1400) and some follow-on risk assessment research by USNRC. The Three Mile Island accident is briefly discussed from a risk viewpoint to illustrate a weakness in current practice. The development of a probabilistic safety goal is recommended with some suggestions on underlying principles. Some ongoing work on risk perception and the draft probabilistic safety goal being reviewed in Canada is described. Some suggestions are offered on further risk assessment research. Finally, some recent U.S. Nuclear Regulatory Commission actions are described.     

核电站工程模拟器用于SGTR事故仿真分析研究

核电站数值反应堆系统(DRS)是基于轻水反应堆瞬态系统分析程序RELAP5的工程模拟器。本工作使用该工具模拟恰希玛(CHASHMA)核电站蒸汽发生器传热管破裂(SGTR)事故，对30min不干预和30min内干预分别进行计算。仿真过程及计算结果验证了数值反应堆系统是进行核电厂仿真和分析的有效工具。     

核电厂选址及环境影响评价应关注的问题

结合近年来我国核电厂选址工作,根据核安全以及环境保护的有关法规要求,重点分析了我国核电厂选址及环境影响评价的特点,并提出了我国核电厂选址及环境影响评价应关注的问题。     

核电运行人才的培养

以秦山第二核电厂为例,介绍了系统性培训方法SAT和运行人才培养体系、培训管理组织机构和责任、核电厂运行人才培养的逐级培训体系、在岗培训制度、运行处的岗位授权制度、人才培养的激励机制等,为核电厂运行人才培养提供参考。