Development of SA-533 Type B CL. 1+SA-240 Type 304L roll-bonded clad steel plate for safety injection tank of CAP1400 nuclear power plant

Aiming to meet the demand of the country＇ s nuclear demonstration project on the CAP1400 nuclear power plant, Baosteel uses the roll-bonding technology and develops the SA-533 Type B CL. 1 ＋ SA-240 Type 304L high-strength and high-toughness clad steel plate with a shear strength of over 310 MPa for the nuclear power plant＇ s safety injection tank. The properties of the quenched and tempered and the simulated post-weld heat treatment states are systematically studied herein through a comprehensive inspection and evaluation of the composition,microstructure,and properties of the clad steel plate. The results show that the bonding interface has high shear strength and that the base metal has high strength and good toughness at low temperatures. Hence, the performance fully meets the technical requirements of the CAP1400 nuclear power plant＇ s safety injection tank in the country＇ s nuclear demonstration project. The roll-bonded clad steel plate can be used to manufacture the safety injection tank of the CAP1400 nuclear power plant.

Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel （PCCV） model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

Development of nuclear power plant real-time engineering simulator

A nuclear power plant real-time engineering simulator was developed based on general-purpose ther- mal-hydraulic system simulation code RELAP5. It mainly consists of three parts:improved thermal-hydraulic system simulation code RELAP5,control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power,was simulated by the en- gineering simulator as an application example. This paper presents structure and main features of the engineering simulator,and application results are shown and discussed.

Study on evaluation system for Chinese nuclear power plants

This paper analyzes the meaning, structure, function and assessment methods of a nuclear power plant evaluation system, and the similarities and differences among various assessment methods. Based on this research an integrated and detailed suggestion is proposed on how to establish and improve internal and external evaluation systems for Chinese NPPs. It includes: to prepare and implement the nuclear power plant operational management program, to build an integrated performance indicator system, to improve the present audit system and conduct the comprehensive evaluation system, to set up and implement the integrated corrective action system, to position precisely the status of operation assessment of nuclear power plants, to conduct the assessment activities on constructing NPP, to initiate the specific assessment in some important areas, to establish industry performance indicator system, to improve the assessment methods, to share the assessment results, to select, cultivate and certify the reviewers, and to enhance international communication and cooperation.

Ageing related events at nuclear power plants

This paper presents main results of a comprehensive study performed by the European Clearinghouse on Operating Experience Feedback of NPP with the support of IRSN (Institut de S?reté Nucléaire et de Radioprotection) and GRS (Gesellschaft für Anlagen und Reaktorsi-cherheit mbH). Physical ageing mechanisms of structure, systems and components that eventually lead to ageing related systems and components failures at nuclear power plants were the main focus of this study. The analysis of ageing related events involved operating experience reported by NPP operators in France, Germany, USA and to the IAEA/NEA International Reporting system, on operating experience for the past 20 years (i.e. 1990-2009). A list of ageing related events was populated. Each ageing related event contained in the list was analyzed and results of analysis were summarized for each commodity group for which the ageing degradation appeared to be a dominant contributor or direct cause. The most common degradation mechanisms/ageing effects for each specific component/commodity group, their risk significance and consequences to the plant performance are described. This paper provides insights into ageing related operating experience as well as recommendations to deal with the physical ageing of nuclear power plant SSC important to safety.

Comparison of neutron energy spectrum unfolding methods and evaluation of rationality criteria

The neutron energy spectrum was measured using a Bonner sphere spectrometer at six locations inside the containment vessel of a nuclear reactor at the Qinshan nuclear power plant. The structures of the neutron spectra obtained by the maximum entropy, iteration, and genetic algorithm methods were consistent with one another and could be interpreted as the spectral superposition of different energy regions. The characteristic parameters of the neutron spectrum, including the fluence rate,average energy, and neutron ambient dose equivalent rate H^(*)(10), were in good agreement among the three methods. In addition, an LB6411 neutron ambient dose equivalent meter was employed to obtain the H^(*)(10) directly for comparison.These findings indicate that neutron spectrum unfolding methods can be used to overcome the problems associated with the response functions of dosimeters to provide more accurate H^(*)(10) values. In this study, the following three evaluation criteria were systematically addressed to ensure the accuracy of the unfolded spectra: count rates of the inverse solutions,neutron spectrum structures, and comparison of key parameters.

Safety of Future NPPs Must Not Be in Conflict with Economics

The conflict between safety and economics requirements is peculiar to the present nuclear power (NP). The main point of the conflict is that for traditional type reactors the increase of requirements to safety of nuclear power plants (NPP) worsens their economical characteristics. This is caused by large potential energy accumulated in reactor coolant. In the presented paper the opportunity and expediency of changeover to reactors with heavy liquid-metal coolants (HLMC) in future NP is grounded. First of all, this refers to lead-bismuth coolant (LBC) mastered in the process of operating nuclear submarines (NS) reactors. The reactor facilities (RFs) of that type cannot cause destruction of defense barriers and make possible deterministic elimination of severe accidents with catastrophic radioactivity release. So it will make possible to eliminate the highlighted conflict and reasons for existence of population’s radiophobia. Lead-bismuth fast reactor SVBR-100 with electric power of 100 MWe is the reactor facility of that type. The effect of accumulated in coolant potential energy on safety and economics is considered. Main specific features of SVBR-100 technology providing a high level of inherent self-protection and passive safety are presented.

模拟事故工况下非能动核电厂安全相关涂层的可靠性测试及评估方法研究

安全相关涂层在非能动核电厂中起着重要的作用,涂层的失效会影响核电厂安全系统的功能执行,影响核安全。国内外核监管机构对于在设计基准事故(DBA)工况下涂层系统的可靠性及评估方法非常重视。文章结合非能动核电厂涂层系统的工程应用,针对其在DBA下的可靠性及评估方法进行了研究。研究表明:在DBA下非能动核电厂安全相关涂层的可靠性要综合考虑涂层的模拟DBA性能、干膜密度、导热性能等。而非能动核电厂安全相关涂层工程应用,则需从涂层的模拟DBA性能、干膜密度、导热性能、涂层碎片(数量、大小、位置和性能等)以及包络涂层碎片后的碎片裕量等角度进行综合评估,以确定在事故工况下涂层的可靠性,不对系统安全产生影响,保证核电厂更安全、高效和经济性运行。

核电工程防造假管理体系建立与优化

近年来,在国内外核电建设过程中发现了个别造假现象,这些造假现象造成了经济损失,带来了质量隐患,引起了舆情风险。本文阐述了核电工程防造假管理体系建立与优化的总体思路,辨识、分析和评估了核电行业的造假风险,针对造假风险制定了防控措施,并探讨了后续的防造假管理体系优化方向,对于提高核电厂工程项目防造假管理能力具有重要意义。

核电厂安全级电气连接器的设计与试验

为实现安全级仪表信号的可靠传输,核电厂通常采用可拆卸的电气连接器连接仪表与电缆以及电缆与电缆。对电气连接器技术进行调研,设计了一种结构简单、性能可靠、操作安装方便、在地震以及严重事故下能有效吸收振动载荷、能够承受更长时间的辐照老化和热老化的安全级电气连接器。依托研制样机开展了功能性能试验和鉴定试验,试验结果表明:安全级电气连接器具有极高的可靠性,能够满足核电厂事故环境下的需求。该连接器可推广于其他恶劣环境条件下的应用领域。

核电厂离线啜吸检测装置抗震性能分析与评定

啜吸检测装置是核电厂燃料操作与贮存系统中的检查设备,分析其在地震载荷下的安全性能具有重要的意义。为了验证设备在地震工况下的安全性和可靠性,校核设备是否满足强度要求,应用有限元软件ABAQUS对啜吸装置进行了建模和抗震计算分析。介绍了设备结构、载荷组合和使用限制、地震反应谱法的分析过程及模型边界条件设定,采用反应谱法计算得到了设备在地震载荷下的响应,并依据相关规范对设备主要部件(筒体、结构件、定位销和锚固螺栓)在静态载荷、地震等多种工况载荷组合作用下的应力进行了强度校核和评定。结果表明,啜吸装置的结构强度满足规范要求,在地震工况载荷作用下能够保证设备结构的完整性并可靠运行。通过抗震计算分析,为啜吸装置优化设计提供了参考依据,对保证核电厂设备的正常可靠运行具有重要意义。

基于风险指引型设备分级的核电厂电动阀预防性维修周期替代技术的研究

本文基于风险指引型设备分级的要求,确定了核电厂电动阀预防性维修替代技术的具体方法和流程,用于保证核电厂安全经济运行并提高电动阀可靠性。

核电厂运行阶段安全文化评价指标体系研究

为培育良好的核电厂运行阶段安全文化,通过分析核电厂运行特征,总结已有的核安全文化评价指标体系和评价模型,构建核电厂运行阶段安全文化评价指标体系,划分为价值观、行为、系统和环境4个层次,并细分出13个二级指标和61个三级指标;在此基础上,考虑到指标之间的非独立性和可能存在的相互影响关系,提出一种基于决策试验和评价实验法(DEMATEL)以及网络层次分析法(ANP)相结合的综合方法,确定指标体系的权重。结果表明:该方法结合调研数据,可得到核安全文化评价指标权重,并甄别出改善核安全文化的关键在于决策层的安全意识、以身作则等指标,为核电厂运行阶段安全文化的培育提供指导。

基于模糊综合评价法的热电厂设备安全风险评估

为定量评价热电厂生产过程中的安全隐患,对热电厂的储煤运煤、锅炉设备、除灰除渣和脱硫脱硝四个子系统中存在的主要危险源进行辨识,并采用模糊综合评价模型进行评估。通过层次分析法确定了热电厂一级主因素和二级子因素的权重并对各因素的危险性大小进行排序;通过计算隶属矩阵得出一级模糊评价矩阵和二级模糊评价矩阵,得出基于专家打分的四个子系统的危险性水平和热电厂的危险性水平,最后计算出该热电厂的安全评价分值为92.7分。所得结论对提高该热电厂的安全水平和后续安全管理具有理论指导作用。

RTM在核级冷水机控制系统标准化设计中的应用研究

随着国内大量核电机组的新建,如何有效实现其中核级冷水机控制系统的设计标准化已成为亟待解决的问题。以某核级冷水机控制系统项目为样本,将需求管理中需求追踪矩阵(RTM)的方法和标准化的常见方法结合,提出一种标准化设计方法。该方法以需求为核心,建立核级冷水机组标准化设计流程,对核级冷水机的设计输入和输出进行分析和建模,实现了冷水机控制系统设计的标准化。通过对该标准化设计在其他实际项目的应用,证明该方法可提高设计质量和效率。该方法的提出与应用为后续其他核电厂主设备的控制系统标准化设计提供了参考。

核电厂反应堆冷却剂系统抗震阻尼比研究

基于2023年国内新建核电厂安全审评中核安全监管部门对抗震物项阻尼比取值提出关注的背景。介绍核电工程中抗震分析阻尼比取值依据,指出标准与工程实践之间存在的差异,以及由此产生的困惑;阐述阻尼比在动态分析中的作用原理;对比核电领域不同标准与导则文件对于机械设备阻尼比的要求,指出当前标准的相关要求对于由多种部件组成的组合设备或系统过于保守;重点对压水堆核电厂反应堆冷却剂系统与设备阻尼比进行研究,给出国内外核电工程实践中该系统与设备的阻尼比取值依据,并针对核电工程实践中组合设备或系统阻尼比取值依据不足的问题提出建议。

核电工程质量诚信风险分级管控及评价体系的建立

对核电工程质量诚信特点和面临的挑战进行了分析,归纳了核电工程质量诚信管控风险分类和动机,结合核电工程管理实践提出了核安全文化、管理体系、组织能力3个维度的质量诚信管控要求和建议,并依照核电工程质量保证分级原则,提出了质量诚信风险分级管控机制及要求,在此基础上构建三维度质量诚信评价模型和五层次评价指标体系,并通过案例对指标应用进行了分析说明,为核电领域及其他行业各级组织提升质量诚信管控水平提供借鉴。

高温气冷堆核电厂主蒸汽管道焊接见证件不合格问题研究和经验反馈

文章主要介绍了高温气冷堆核电厂主蒸汽管道焊接见证件冲击试验不合格问题中,营运单位对问题的处理和原因分析过程以及核安全监管部门现场监督情况。本文的目的是通过对事件经过及现场监督活动的梳理总结,得到一种可以借鉴的类似问题处理经验,促进相关单位能够更好地在法律法规和相关标准要求下开展核电厂建造活动,进一步提高营运单位全员核安全文化意识,加强与监管部门的沟通,同时做好经验反馈工作。