核电厂1E级安全壳内电动机鉴定试验相关问题的探讨

核电厂1E级安全壳内用电动机作为在正常工况下和设计基准事件期间及之后向核电厂安全系统设备提供动力的重要设备之一,必须按照相关的标准制定鉴定大纲和程序,进行鉴定试验。对1E级安全壳内电动机的鉴定过程和鉴定文件进行审查是核安全设备审查的重要内容之一。本文结合实际审评经验,探讨了电动机的加速热老化试验和LOCA试验中相关问题,提出了审评者的观点。     

我国核电厂抗震设计反应谱和RG 1.60设计反应谱的比较分析

设计反应谱对评价核电厂在地震作用下的安全性极为重要。本文从统计核电厂抗震设计标准反应谱时选取的强震数据及统计方法两个方面,分析比较了美国RG 1.60设计反应谱和我国核电厂抗震设计规范反应谱的异同。通过对比分析,深入理解核电厂抗震设计反应谱的提出需考虑的关键因素,为核电厂抗震设计和审评工作提供参考。     

EPR核电厂设备安全分级审评和评述

介绍了EPR核电厂设备安全分级方法,阐述了台山核电厂审评中的主要观点,对比M310机组,评述了EPR核电厂设备安全分级问题审评工作,对今后EPR机组的安全审评提出了建议。     

基于抗震鉴定试验的开关柜的抗震裕度分析

核电厂发生超过设计基准地震后,需要进行抗震裕度分析以便于识别核电厂的薄弱环节。本文利用高置信度低概率失效来量化设备的抗震裕度,采用保守的确定论失效裕度和易损性分析两种方法,计算了核电厂设备的高置信度低概率失效,梳理了两种方法的计算步骤,明确了计算过程中关键参数的取值范围。利用两种方法计算基于抗震鉴定试验的开关柜的高置信度低概率失效。     

核电厂气载流出物大气弥散因子不同计算方法比较研究

当前,基于核电厂气载放射性流出物环境影响的评价模式种类较多,然而这些模式及其开发的程序多用于评价过程,专门针对环境后果影响审评的应用较少。本文从分析核电厂气载放射性流出物大气弥散的相关导则标准以及基于此开发的相应程序入手,针对典型核素,采用C-Airdos,ROULEA-2.0和Fluidyn-PANACHE等程序,以我国自主设计的CAP1400压水堆重大专项示范工程石岛湾核电厂所在地环境数据为基准,开展正常运行工况下大气弥散因子模拟计算,与石岛湾核电厂环境影响评价报告所采用的Y3001程序计算结果进行比较分析,推荐适用于核电厂正常运行工况下气载流出物环境影响的审评模式及程序,为我国核电厂环境影响审评提供有益的技术参考。     

先进轻水反应堆业主要求文件（ALWR—URD）中对核电厂抗震设计要求

“先进轻水反应堆业主要求文件(ALWR—URD)”的宗旨是明确美国电力公司对先进轻水堆核电厂的要求。对URD中关于核电厂抗震设计，特别是对核电厂构筑物、系统和设备的抗震分类、取消OBE地震后的抗震设计要求等方面提出的修正意见．以及对核电厂抗震裕度和地震风险评价提出的初步要求，本文作了全面的总结，在关键处浅讨了看法。供新核电厂设计中参考。     

核电厂电仪设备延寿再鉴定试验研究

在核电厂电气仪表设备（简称电仪设备）环境鉴定研究成果的基础上,开展核电厂电仪设备延寿再鉴定分析和试验研究。以秦山第一核电厂DDG-1型电气贯穿件（EPA）为研究对象,根据运行实际制定了再鉴定试验研究的遵循原则,在此原则下结合分析法确定了试验方案和试验项目序列以及EPA修复依据和方案,并在此基础上开展再鉴定试验研究。适当修复后的DDG-1型EPA按试验大纲依次通过了设备性能随时间变化的试验、抗震试验、设计基准事故（DBA）条件下热力学试验和DBA后极限电性能试验,试验后状态完好,表明该DDG-1型EPA经适当修复后能够完成继续延寿20 a的预期目标,可为核电厂其他电仪设备再鉴定试验研究提供指导和借鉴。      

研究实验反应堆的核设备审评

介绍了研究实验性反应堆的核设备审评目的和主要求。核设备失发级原则和相应设计制造要求。核设备的应力分析的抗震要求，以及一回路设备的断裂韧性的特殊要求，指出审评中发现的问题，并对如何解决这些问题给出了一些建议。     

核电厂管线上设备地震加速度响应放大系数评估方法与应用

为了方便快捷地确定核电厂管线上设备的地震载荷,基于弹簧-质量模型、管线与安装于管线上的刚性设备的关系,提出了评估管线对设备的地震加速度响应放大系数的方法,以及该评估方法在确定设备抗震鉴定地震输入方面的应用方法,并通过有限元模型对该评估方法和应用予以了验证。结果表明,该评估方法在确定管线上设备抗震鉴定地震输入方面提供了易于操作又不至于太过保守而降低经济性的可行的应用办法。      

浅析美国核电厂执照更新的物项筛选原则

我国部分核电厂正在开展核电厂运行许可证延续论证工作。延续论证首先需要确定纳入审查范围的系统、构筑物和设备,并需要对这些系统、构筑物和设备进行评价,以确保核电厂的安全性。本文解读了美国法规对核电厂执照更新中物项筛选原则的要求,并对其进行分析和研究,为我国核电厂运行许可证有效期限延续申请和相关审评工作提供了参考依据和支持。     

Recent extensions to seismic margin review methodology

Two seismic margin review methodologies — one by USNRC and the other by EPRI — have been developed in the last four years. The focus is on assessing the capability of existing nuclear power plants to withstand earthquakes larger than the design basis earthquakes. The methods restrict the analysis to a selected few systems and components using the insights from past seismic PRAs, seismic analysis and qualification results, and earthquake experience data. The objective of this paper is to describe recent and on-going studies in extending the NRC seismic margin review methodology. Specifically, three topics are discussed: (1) extension of the HCLPF capacity to analyse radiological releases and importance of human factors and non-seismic failures; (2) importance of BWR plant systems and functions to seismic margins; and (3) extensions of seismic margin review results to obtain seismic risk estimates.     

Seismic resistant design of safety class structures and equipment

Some commonly encountered problems in the seismic resistant design of nuclear power plant facilities are discussed. The topics included here are ground input motions, local geology versus source mechanism and travel path, three components inputs, torsional responses, floor response spectra, seismic resistant design of heavy equipment, the application of component mode synthesis technique, seismic resistant design of piping systems, equipment qualification by testing, the effects of close modes, underground pipe design, and soil structure interaction.     

基于有限元模型验证的控制室盘台抗震鉴定

论文以国内某新建核电站控制室盘台抗震鉴定为例,阐述了基于有限元模型验证的盘台抗震鉴定方法。通过样机试验和模型验证分析,将盘台结构设计与有限元分析进行了有机结合,同时在盘台整体有限元分析验证过程中引入了修正因子,保证了盘台的抗震性能,并使其具有一定的安全裕度。     

核电厂1E级充电器及逆变器设备鉴定方案的研究

基于IEEE、IEC、RCC-E等标准及文献,阐述了1E级电气设备的质量鉴定方法,进而结合国内外设备鉴定的实践经验,制定了1E级充电器、逆变器的鉴定方案,对元器件评估、性能及应力试验、EMC试验、抗地震试验、软件鉴定等环节进行重点剖析。可为1E级充电器、逆变器设备国产化过程中的质量鉴定提供参考。     

Seismic qualification of expansion anchors to Canadian nuclear standards

A large number of unique seismic qualification tests were performed on different types of expansion anchors to determine their seismic capabilities. These tests followed the requirements of the Canadian CSA N287.2 as well as the ASTM E 488 Standards. Expansion anchors ranging in diameter from 1/4 to 11/4 in. were subjected to this seismic qualification testing. Strategically designed and standardised concrete blocks were manufactured to receive each expansion anchor. The anchors were installed according to the anchor manufacturer instructions and the installation was consistent with construction practice. Each anchor type and size was subjected to four different types of tests covering static, dynamic, tension and shear loading. The dynamic cyclic loads applied represented a strong earthquake whose spectrum envelops that of the Design Basis Earthquake (DBE). The seismic testing qualification program involved 300 tests. Results of the testing program are obtained for each type of anchor. Results from these tests confirm that these types of expansion anchors have maintained their tension (pull-out) and shear capabilities following a severe seismic shaking.     

Standardized seismic design spectra for nuclear plant equipment

Standardized seismic design spectra were developed for the nuclear plant equipment. Such standardized spectra will reduce unnecessary spectrum computation for the individual plants as well as expedite the selection, purchase and qualification of equipment. Most attractive of all, standard equipment such as electric components can be conservatively qualified just once and then used by all plants located in the same seismic zones. Presented here are idealized design spectrum peak envelopes for different types of structure materials and for both the SSE and OBE conditions. Procedures are also provided for the application of the design spectrum peak envelopes to the seismic qualification of equipment either by dynamic analysis or testing.     

反应堆堆顶和控制棒驱动机构整体抗震分析

通过建立反应堆堆顶和控制棒驱动机构(CRDM)的整体三维有限元简化抗震分析模型,对反应堆堆顶和CRDM进行了三维非线性抗震分析。该分析模型可真实反映CRDM顶部抗震支承板相互之间、抗震支承板与抗震支承环之间的碰撞作用,以及抗震拉杆的拉压非线性特性,进而获得堆顶和CRDM结构各位置的精确分析结果。通过分析,得到反应堆堆顶和CRDM各部位的载荷,可为部件的应力分析提供必要的地震载荷,为CRDM抗震鉴定试验提供加速度时程等输入数据。     

高温气冷堆蓄电池组地震易损性研究

为验证核电厂发生地震外部事件时的电力安全,需要对蓄电池组进行抗震鉴定试验。本文以高温气冷堆(HTR)核电厂安全级蓄电池组为研究对象、以安全级蓄电池组抗震鉴定试验数据和工程经验为基础,通过识别、量化蓄电池组的地震易损性变量,并应用基于试验的易损性分析法推导出地震易损性曲线和高置信度低失效概率(HCLPF)抗震能力。研究结果表明,安全级蓄电池组的抗震能力远高于核电厂设计基准地震动需求。