核电厂电气贯穿件设备延寿再鉴定方法研究

以秦山一期核电厂电气贯穿件(EPA)为实施对象,基于已有设备设计、制造、鉴定和实际运行数据,并结合国内外核电厂的设备鉴定、老化延寿管理等标准规范和最新研究成果,开展了核电厂设备延寿再鉴定方法研究,同时结合目前国内核电厂类似EPA等1E级核安全设备老化管理的现状,提出了设备老化管理建议。     

核电厂安全级电气设备热老化试验方法概述解析

核安全级电气设备执行安全功能的能力可能受到环境条件和运行条件随时间变化的影响,因此核安全级电气设备的鉴定技术是保障核电厂经济运行和延长现有核电厂运行寿命的必要手段,核安全级电气设备的加速热老化试验作为鉴定试验的一个过程至关重要,本文主要介绍了目前常用的加速热老化试验方法,包括直接选取活化能获得的热寿命评定、基于热寿命曲...     

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。     

1E级电缆概率可靠性剩余寿命预测方法

1E级电缆是影响核电厂安全、可靠运行的重要部件之一。由于受多种老化降质因素的影响,在整个电厂寿期内又难于进行定期维修更换,随运行时间延长会发生老化或降质,需要有效的状态监测方法来进行监督。压入模量因其现场无损特性和与断裂伸长率良好的对应性,能够有效的对电缆老化进行跟踪,是一种有效的电缆状态监测方法。为开发一种建立1E级电缆压入模量与断裂伸长率关系模型和使用概率可靠性分析预测其剩余寿命的综合方法,对国内核电厂某种典型的1E级电缆在399 K-360 Gy/h下进行了热-辐照加速老化,测试了护套和绝缘的断裂伸长率及护套的压入模量。建立了电缆护套压入模量与绝缘断裂伸长率关系模型,分析了其在不同工作温度下的可靠性,开发了基于概率可靠性分析的1E级电缆剩余寿命预测方法。结果显示,累积辐照剂量为375 kGy时,在313 K和338 K的工作温度下,电缆预期剩余寿命分别为60年和35年。该方法可推广到电缆其他老化机制剩余寿命预测上。     

应用于核电厂老化PSA的SSC筛选方法研究

核电厂运行许可证延续必须考虑其延寿期内的核安全问题,确保核电机组在延期运行期间的核安全水平不低于原设计寿期内的核安全水平。可应用PSA技术对许可证延续期间的核电厂建立老化PSA模型,从而评估SSC老化对核电厂整体安全的影响,验证其仍可满足原设计标准。基于此提出了应用于核电厂老化PSA的SSC筛选分析方法,通过考虑趋势分析,老化失效模式与影响分析,风险重要度分析,在三种分析方法基础上建立核电厂SSC筛选的决策矩阵,为选择易老化且安全重要的部件建立了可行的方法。该项工作也为核电厂在许可证延续阶段的风险指引型管理奠定技术基础。     

国内核电厂安全壳地坑滤网设计改进考虑

在冷却剂丧失事故（LOCA）工况下核电厂安全壳地坑滤网堵塞问题已是核能界广泛关注的核安全问题,国内核安全监管部门和核电厂营运单位正积极推动该问题的解决。本文介绍了国内核电厂安全壳地坑滤网设计改进工作的进展情况,从审评人员的角度说明了对解决该问题所持的态度及相应的监管要求,并阐述对国内相关核电厂逐步开展该项工作的总体设想。     

CPR1000余热排出泵地震工况可运行性评价准则初步研究

核安全级泵地震工况的可运行性直接影响核电厂安全,需通过鉴定验证其地震工况下的功能。文中以CPR1000余热排出泵为研究对象,利用故障模式及影响分析(FMEA)方法进行地震工况下泵的故障模式及影响分析,筛选出影响泵可运行性的薄弱环节,并基于零部件的功能和运行要求,提出地震工况下泵组可运行性的评价准则。     

核电厂1E级电缆的核安全审查

阐明对核电厂1E级电缆核安全审查的基本要求.     

秦山核电厂反应堆压力容器辐照脆化时限老化分析

反应堆压力容器是核电厂最重要设备之一,其辐照脆化状态决定了核电厂的实际运行寿命。通过借鉴国外反应堆压力容器安全评估方法,开发出一套反应堆压力容器辐照脆化时限老化分析（TLAA）的方法。该方法从上平台能量、反应堆运行压力-温度曲线及承压热冲击3个方面评价压力容器材料在正常工况和事故工况下的安全裕度。采用该方法在秦山核电厂运行许可证延续（OLE）项目中对反应堆压力容器进行了辐照脆化TLAA安全评估,其评估方法和评估结论到得国家核安全监管局的认可,为秦山核电厂延寿20 a奠定了基础。     

不受环境鉴定要求约束的核电厂低压电缆老化管理

本文基于对电缆老化应力、局部恶劣环境识别、电缆状态监测技术以及国际核电厂电缆管理经验的研究和分析,给出核电厂不受环境鉴定要求约束的低压电缆的老化管理方法,可为国内运行核电厂非环境鉴定要求的低压电缆的老化管理提供支持和参考。     

Loss-of-coolant accident (LOCA) testing of aged cables with application to nuclear plant life extension

Sandia National Laboratories is conducting long-term aging research on representative samples of U.S. nuclear power plant Class 1E cables. The objectives of this program are to determine the suitability of these cables for extended life (beyond 40-year design basis) and to assess various cable condition monitoring (CM) techniques for predicting remaining cable life. Twelve different cable products have been artifically aged at relatively mild exposure conditions in three test chambers to nominal equivalent lifetimes of 20, 40, and 60 years based on Arrhenius acceleration theory. After aging, the cables in each chamber were exposed to a sequential accident profile consisting of 1100 kGy of high dose rate gamma irradiation followed by a simulated design basis loss-of-coolant accident (LOCA) steam exposure. This paper presents selected results of the LOCA testing. Although some of the cables experienced electrical failures, the results of these tests indicate a good life extension potential for a number of popular U.S. cable products. Results of the CM techniques are still under evaluation and are not reported here.     

Condition monitoring methods applied to class 1e cables

Sandia National Laboratories is currently conducting long-term aging research on representative samples of nuclear power plant Class lE cables. The objectives of this program are to determine the suitability of these cables for extended life (beyond 40-year design basis) and to assess various cable condition monitoring (CM) techniques for predicting remaining cable life. The cables are being aged for long times at relatively mild exposure conditions with various CM techniques being employed during the aging process. Following the aging process, the cables will be exposed to a sequential accident profile consisting of high dose rate irradiation followed by a simulated design basis loss-of-coolant accident (LOCA) steam exposure.This paper covers two aspects of the research program: the laboratory measurements and on-line electrical measurement techniques that have been developed and are being performed and initial data that has been generated from the on-line measurements. The electrical measurement results presented include insulation resistance, polarization index, capacitance, and dissipation factor from a few of the cable samples tested.     

The importance of oxygen in LOCA simulation tests

Experiments were performed to survey the effects on material degradation of both aging conditions and the oxygen concentration during a Loss-of-Coolant Accident (LOCA)-simulation. Changes for a number of commercial materials commonly used as electric cable jackets and insulations in nuclear power plant applications were monitored in terms of weight, mechanical properties, solubility measurements and infrared spectroscopy. For a number of these materials (an EPR insulation, a chloroprene jacket and a PVC jacket), the concentration of oxygen during LOCA simulation was found to be an important parameter. For the first two materials, more degradation occurred when oxygen was present; for PVC, substantially increased swelling occurred as the oxygen concentration was lowered. These results indicate that a realistic LOCA-simulation test should include an overpressure of oxygen gas representative of the postulated oxygen concentration expected during a predicted LOCA. Conditions of accelerated aging exposure prior to the LOCA simulation were also found to have a very substantial influence. In particular, for a number of the materials, lowering the radiation dose rate used for aging led to enhanced degradation after both the aging and the LOCA simulation.     

基于多分类SVM的船用核动力装置主回路系统破口特征诊断技术研究

失水事故发生时,在事故初期判断出事故类型对操纵员安全操作意义重大,为此提出一种基于监控参数的失水事故类型判断方法。该方法根据事故发生后13 s内监控参数的变化速率与破口类型的对应关系,提取故障征兆,建立事故判断模型,并根据建立的模型使用支持向量机分类的方法进行破口事故类型判断。试验结果表明,该方法在事故发生初期可准确、有效地判断出典型失水事故的破口尺寸和相对位置。     

核电厂大LOCA始发严重事故下氢气源项的敏感性分析

采用MELCOR程序,以600MW级核电厂为研究对象,在以大破口失水事故为始发事件的严重事故中,针对不同的破口尺寸及破口位置对堆芯内锆-水反应及堆腔内熔融堆芯与堆腔混凝土之间的相互作用(MCCI)中氢气源项的影响进行敏感性分析.结果表明,在大破口始发的严重事故中,不同的破口尺寸对氢气源项的影响不大;而在破口尺寸相同的情况下,破口发生在主管道热段时,产氢速率的峰值最大;破口发生在主管道冷段时,累积的总产氢量最大.     

华龙一号安全壳热工响应确定论现实方法研究

失水事故（LOCA）是压水堆核电厂的一种典型设计基准事故,该事故后的安全壳热工响应过程,尤其是安全壳压力峰值直接影响安全壳结构的完整性。本文采用确定论现实方法（DRM）对华龙一号核电厂LOCA质能释放与安全壳热工响应进行分析研究。对关键参数进行敏感性分析及统计计算,并建立DRM惩罚模型。计算结果表明,DRM惩罚模型的计算结果始终高于95%置信水平下、95%概率下的统计计算值,DRM惩罚模型是保守的。DRM方法对于华龙一号核电厂的LOCA质能释放与安全壳热工响应分析是适用的。     

In search of extendable conditions for cable environmental qualification in nuclear power plants

The environmental qualification (EQ) for cable insulators in nuclear power plants (NPPs) has been developed on the basis of the design basis accident (DBA) to prevent reactor core damage. However, the latest safety principles require extending the design concept to prepare the utilized equipment for scenarios after core damage. Thus, we propose a modification to the EQ for cables connecting utilized equipment at design extension conditions. This paper surveys all electrical components for accident management in boiling water reactor-4 (BWR-4), and identifies their connecting cables’ functional category as low-voltage power, instrumentation, and control cables. The EQ temperature profile of these cables during the incident phase was addressed for extension. This required postulating maximum temperature environments according to accident scenarios, knowledge of cable integrity degradation, and their current evaluation by the EQ. To evaluate whether these environments are suitable stressors, heat testing was conducted on flame-retardant ethylene propylene rubber (FR-EPR)-insulated cables. On the basis of those results, we suggest a maximum primary peak temperature of the EQ temperature profile of 250 °C. We also suggest increasing the primary peak period of the EQ temperature profile to 48 h without experiment, on the basis of inherent excessive margin for mechanical integrity during the ageing phase.     

燃料热导率降级对CAP1000大破口失水事故的影响分析

在高燃耗情况下,燃料芯块的热导率随燃耗降低,该现象被称之为热导率降级(TCD)现象。TCD现象影响失水事故(LOCA)前稳态工况的燃料平均温度和燃料储能,进而影响大破口LOCA过程中的包壳峰值温度(PCT)。本研究采用大破口LOCA分析程序WCOBRA/TRAC对CAP1000冷段双端剪切断裂事故进行了不同燃耗的敏感性分析,并获得了不同工况下的PCT。分析中采用美国核燃料研究所(NFI)修正的TCD模型对降级后的燃料热导率进行模拟,同时考虑了燃耗大于30GW·d/tU后FQ和FΔh峰值因子的降低。敏感性分析表明,考虑TCD和峰值因子降低的影响,PCT极限工况不再出现在低燃耗区间,而出现在燃耗为29GW·d/tU附近。与其他燃耗水平相比,该燃耗点的PCT第1峰值和第2峰值均处于最高水平。本研究结果可为高燃耗情况下非能动电厂大破口LOCA的分析评估提供参考。     

Experiments on development, spreading and detection of cable fires

Tests were carried out to investigate the development and spreading of fires in electrically overloaded cables, and the fire detection capabilities of various smoke detectors. Three tests were performed on horizontal cable trays having different cable arrangements. The development and propagation of the fire between cables and cable trays was observed and its influence on the functional capability of neighbouring cables noted. The development and spreading of smoke was measured together with the actuation of the smoke detector systems.Test results showed that each of the smoke detectors was able to detect a smouldering cable and actuate alarms well before the cable being tested actually caught fire, and well before adjacent cables or cables in neighbouring trays, would have lost their ability to perform their normal electrical functions. Such test results can provide useful input to risk studies involving the assessment of cable fires.     

Experimental results on the breakaway oxidation of the E110 cladding alloy under high-temperature isothermal conditions

In order to evaluate the effect of zirconium breakaway oxidation on the behaviour of nuclear fuel in LOCA (Loss-of-coolant accident) conditions, high temperature (800–1200 °C) oxidation tests with E110 type cladding were performed in steam atmosphere. The onset time of breakaway oxidation was detected using an online method based on hydrogen release. The experimental results showed that the breakaway oxidation starts not earlier than 5 min after the start of the oxidation that is longer than the duration of dry phase in a design basis accident LOCA. The experiments give preliminary indication that the breakaway oxidation does not play role in design basis accident LOCA conditions with the E110 alloy. This is to be confirmed by transient oxidation tests.