Reliability data update method for emergency diesel generator of Daya Bay Nuclear Power Plant

In the field of Living Probabilistic Safety Assessment (LPSA) the reliability data updating is an important factor. In risk analysis equipment failure data is needed to estimate the frequencies of events contributing to risk posed by a facility. Five years data of emergency diesel generator (EDG) of Daya Bay Nuclear Power Plant (NPP) has been studied in this paper. The data updating process has been done by using two methods, i.e., the classical method and Bayesian method. The aim of using these methods is to calculate the operational failure rate (λ) and demand failure probability (p). The results show that the operational failure rate is 1.7E?3 per hour and the demand failure probability is 2.4E?2 demand per day for Daya Bay NPP. By comparing the results obtain from classical and Bayesian methods with EDF (Electric De France) it is concluded that the design and construction of Daya Bay NPP is very different than EDF therefore the reliability parameters used in Daya Bay NPP is based on the classical method.     

Analysis of accidents leading to local flow stagnation in parallel fuel channels of RBMK-type reactors

The RBMK-type nuclear power reactors, still operating in Russia, are graphite-moderated with vertical fuel channels, using low-enriched nuclear fuel. The main challenge, which leads to the overheating of the fuel assemblies, fuel channels and other core components in channel type nuclear reactors, is a misbalance between heat generation in core structures and heat sink, which can appear due to the loss of coolant accident. In this accidental case, the emergency core cooling system ensures the core cooling. In RBMK-type reactors this system consists of hydro-accumulators and a number of pumps, taking water from large water reservoirs. This equipment injects water into the fuel channels through the group distribution headers at high pressure. However, the direct supply of cold water from emergency core cooling system into fuel channels is possible only if check valves on group distribution headers are closed properly. If these check valves failed, the part of water would be lost through the break, the flow stagnation in channels could occur, which might lead to overheating of fuel assemblies in the fuel channels. This paper presents the results of deterministic safety analysis, performed using system thermal hydraulic code RELAP5. Using this code the reactor cooling system of RBMK-1500 was modelled and analyses of loss of coolant accidents with failure of few check valves in group distribution headers were performed. The results of the calculations are used for the development of symptom-based emergency operating procedures for RBMK-1500. The basic principles that describe the complex distribution of water flows in vertical forced circulation circuit with parallel fuel channels can be adjusted for the RBMK-1000 reactors, still operating in Russia.     

基于PSA的压水堆LBLOCA不确定性分析

为了结合确定论与概率论分析开展更加真实的核反应堆事故工况安全分析,提出了一种结合概率安全分析（PSA）和最佳估算加不确定性（BEPU）分析的方法,并以典型三环路压水堆冷管段双端断裂大破口失水事故（LBLOCA）的极限事故为对象,首先基于PSA开展了应急堆芯冷却系统的事故失效分析,而后结合BEPU分析评估了事件树中各事故序列的包壳峰值温度（PCT）分布及条件堆芯损坏概率（CCDP）,最终确定了压水堆在该事故工况中的堆芯损坏频率（CDF）。分析结果表明,压水堆在冷管段双端断裂工况中应急堆芯冷却系统能够保证反应堆的安全,且一列低压安注系统足以排出堆芯余热及保证反应堆安全。      

XAPR中破口失水事故下堆芯自然循环冷却能力的功能可靠性研究

功能失效是导致自然循环系统运行失效的重要因素,需在其可靠性分析中予以考虑。针对多维不确定性参数及小功能失效概率问题,提出了一种将改进响应面法及重要抽样子集模拟法相结合的功能可靠性分析方法。以西安脉冲堆（XAPR）堆池水自然循环冷却为例,结合中破口失水事故,考虑输入参数及模型的不确定性,对其进行了功能可靠性评估和灵敏度分析。结果表明：XAPR堆芯自然循环功能失效概率为3.796×10^-3,需充分考虑系统功能的可靠性。本文方法具有较高的计算效率,同时又能保证很高的计算精度,对XAPR堆芯自然循环非线性功能函数具有很强的适应性。     

Underground nuclear power plants with surface turbine generators

This study has produced conceptual designs for light-water reactor underground nuclear power plants with portions of the plant at the surface. The investment cost penalty for underground excavation and cavity lining less interest during construction and escalation attributed to the underground portion of the plant at a favorable geologic site was estimated at between 3 and 4% above a comparable surface plant. The power plant turbine-generator was assumed to be at the surface and the reactor placed underground. Other equipment was located underground on the basis of a hazards analysis or by a functional relationship to the reactor. A guideline was adopted that the nuclear steam supply system and other principal subsystems should not require major redesign. Although some component relocation was allowable, it was permitted only if system performance and component operation were thought to be essentially unaffected. Plant size was stipulated as a single unit of 1000 MW(e) net which was not varied to determine an optimum size or number of units. Condenser cooling methods considered included natural convection wet cooling towers as well as once-through condenser cooling.In both the boiling-water (BWR) and pressurized-water (PWR) plant designs, four underground chambers are proposed to efficiently house the underground equipment. One of the four chambers is needed for the reactor and steam supply system. It is conceivable, if only reactor hazard protection is desired, that the equipment in two of the chambers (the nuclear auxiliaries and relay and switching) could be installed within a surface building. The location of the emergency core cooling system (ECCS) components as presently configured in surface plants relative to the nuclear steam supply system is critical to the satisfactory operation of the ECCS. In some situations the ECCS must draw water from the reactor area. The need for a positive head from this area to the appropriate pumps implies the pumps must be placed lower in elevation than the source of water. Consequently, it was appropriate to locate these ECCS components in a separate underground chamber near the reactor chamber.     

EAST低温系统的故障树分析

EAST是世界上进行聚变研究的先进超导托卡马克实验装置.低温冷却系统是EAST的主要子系统之一,担负着纵向和极向场线圈、纵场线圈盒、冷屏和电流引线的冷却功能,需保证其能长期稳定运行.本文运用故障树分析法来评价低温系统的可靠性,使用自主开发的软件RiskA建立低温系统的故障树模型并进行定量计算,得到系统的失效概率和部件的重要度,并提出改进建议,为系统的优化设计和提高可靠性提供参考.     

Reliability assessment of emergency exhaust system in a pool-type research reactor

The reliability of an extract system in a swimming-pool-type research reactor has been assessed. A global fault-tree analysis technique has been utilized. The basic event reliability data is based on both generic and reactor specific informations.The unavailability of the extract system is quantified in terms of the unavailability of the various functional requirements of the system. The unavailability is expressed as the probability of failure on demand. The computer system unavailability is determined from the minimal cutsets of the system. It is found that only three events have a major contribution to the top event, i.e., failures of compressed air supply, electric power supply and solenoid valve. A sensitivity analysis is performed to show the effects of variations in the data values of the dominant cutsets. An uncertainty analysis was also performed on the fault tree. The evaluations show that the reactor extract system lacks diversity and redundancy in most of its components. It is tolerant of most minor degradations, as these are taken care of by the operating policies and procedures. However, it can not tolerate common cause failures, e.g., simultaneous compressed air and electric power supply failure. Based upon the results obtained some recommendations are made.     

大型非能动压水堆核电厂安全注射系统旋启式止回阀阻力变化对长期堆芯冷却影响分析

大型非能动压水堆核电厂在发生失水事故(LOCA)后的长期堆芯冷却阶段依靠重力向堆芯注入应急冷却水,其注射管线上设置的旋启式止回阀的阻力可随流量变化,管线的阻力可能将非预期地增加。根据旋启式止回阀阻力特性,为失水事故最佳估算系统分析程序添加相应的计算功能,对压力容器直接注射(DVI)管线双端断裂事故后长期堆芯冷却工况进行了计算分析。结果表明:安全注射管线上旋启式止回阀阻力变化对大型非能动压水堆核电厂LOCA后长期冷却的影响较小;在安全裕量不足的情况下,旋启式止回阀的阻力特性将影响到非能动注射管线的安全注射功能的执行。     

严重事故下核电站安全壳内氢气分布及控制分析

使用安全壳分析程序CONTAIN计算分析了百万千瓦级压水堆核电站严重事故下安全壳内的氢气浓度分布.分别对一回路冷段大破口失水(LB-LOCA)叠加应急堆芯冷却系统(ECCS)失效(不包括非能动的安注箱)事故和全厂断电(SBO)叠加汽轮机驱动的应急给水泵失效事故两个严重事故序列进行了计算.计算结果表明,不同严重事故下,安全壳各隔间对氢气控制系统的要求不同.氢气控制系统的设计必须满足不同事故下的法规要求,提高电站的安全性.     

用RETRAN程序进行乏燃料元件贮存水池的热工水力安全分析

开发了对核电厂乏燃料贮存水池进行热工水力分析的ＲＥＴＲＡＮ模型,按照最大热功率工况,即在乏燃料贮存水池中装满乏燃料组件（其中包括换料期间刚卸出的全堆芯燃料组件）的条件下用ＲＥＴＲＡＮ模型来评估乏燃料贮存水池冷却系统的冷却能力,并进行了几个假想方案的瞬态计算和校对计算。利用ＲＥＴＲＡＮ模型来评估乏燃料贮存水池稳态和瞬态的热工水力安全分析既方便,又精确,还可用于申请许可证的计算和估算水池的温度分布。     

地震下非能动堆芯冷却系统可靠性分析

地震情况下核电站非能动堆芯冷却系统（PXS）能否可靠运行对核电站的安全性有着重要影响。本文采用故障树方法分析计算了PXS各部件在峰值地面加速度（PGA）为0.5g、1.5g、2.5g情况下的失效概率以及各部件对系统失效的贡献,并与《AP1000概率安全分析报告》中的抗震裕量分析（SMA）方法的结果进行比较,分析部件的抗震能力。结果表明：本文方法计算的条件失效概率和各部件对系统失效的贡献与SMA方法的结果基本相符。本文方法可为AP1000等非能动核电站的安全分析提供参考。     

Use of the local Pu-239 concentration as an indicator of burnup spectral history in DYN3D

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density.     

基于最优化线抽样法的XAPR自然循环功能失效研究与分析

针对多维不确定性参数、小失效概率的功能可靠性分析,提出了一种优化线抽样的可靠性分析方法。该方法采用遗传算法求解约束条件的优化模型来寻求最优化重要方向,进而得到失效概率的高效估计。以西安脉冲堆（XAPR）自然循环冷却堆芯能力的可靠性评价为例,考虑模型与输入参数的不确定性,对中破口失水事故下的自然循环功能失效概率进行了量化分析。结果表明：与其他概率评估方法相比,本文方法具有很高的计算效率,同时又能保证很好的计算精度;对隐式非线性的功能可靠性分析是有效可行的,具有很强的适应性。     

Seismic failure probability evaluation of redundant fast breeder reactor piping system by probabilistic structural response analysis

The seismic failure probability and the correlation coefficient of the multiple failure mode of the heat transport system of a three-loop fast breeder reactor have been evaluated based on a probabilistic structural response analysis. It has been found that the most probable failure mode of the heat transport system has less impact on the core cooling capability than other modes. The correlation coefficient of the heat transport system loops is approximately 0.9. It is found that the correlation comes from the common structural properties rather than the common seismic input. The present approach is useful for quantifying the correlation coefficient and the seismic fragility of the redundant component failure that is used in the systems analysis.     

危冷系统对蒸汽发生器传热管破损事故缓解能力分析

针对蒸汽发生器传热管破损事故后果严重,危冷系统投入对事故缓解能力认识不清的问题,为提高危冷系统对事故缓解能力的认识,增强传热管破损事故处置能力,利用MELCOR程序建立了破损安全分析模型。通过计算对比分析了危冷系统投入与否对事故后果的影响,并比较了危冷系统对不同尺寸传热管破损事故的缓解能力。经仿真分析,明确了危冷系统对传热管破损事故的缓解能力,对提高运行人员事故处置能力及保证反应堆运行安全有重要意义。      

数字化安全级DCS紧急停堆系统共因失效分析

以2oo3架构数字化安全级分布式控制系统（DCS）紧急停堆系统为研究对象,采用Markov方法对其建立可靠性模型,分别计算并对比了考虑共因失效和不考虑共因失效2种情况下紧急停堆系统的拒动概率,同时对系统拒动概率相对于共因失效因子变化的敏感性进行了重点分析。结果表明,拒动概率随着共因失效因子的增加而变大,因此,在系统设计中需采取有效措施对冗余系统的共因失效进行控制,降低共因失效因子,从而提高紧急停堆系统的可靠性。      

重要厂用水系统地震情况下可靠性分析

重要厂用水系统是核电厂重要的安全系统之一,其失效概率通常由系统可靠性分析获得。而地震情况下设备的失效概率是地震动峰值加速度的函数,且地震的发生又具有随机性,目前概率安全评价中传统的故障树分析方法对此种情况缺乏足够的处理能力。本文采用蒙特卡罗模拟方法解决条件概率的问题,针对地震情况系统可靠性分析,提出了评价模型,并对核电厂重要厂用水系统进行了分析计算,得到地震情况下重要厂用水系统的年失效概率为1.46×10^-4。计算结果与设备抗震性能数据符合,验证了分析模型的合理性。     

Methodology for reliability allocation based on fault tree analysis and dualistic contrast

Reliability allocation is a difficult multi-objective optimization problem.This paper presents a methodology for reliability allocation that can be applied to determine the reliability characteristics of reactor systems or subsystems.The dualistic contrast,known as one of the most powerful tools for optimization problems,is applied to the reliability allocation model of a typical system in this article.And the fault tree analysis,deemed to be one of the effective methods of reliability analysis,is also adopted.Thus a failure rate allocation model based on the fault tree analysis and dualistic contrast is achieved.An application on the emergency diesel generator in the nuclear power plant is given to illustrate the proposed method.     

确定论与概率论相结合的小型模块化压水堆应急堆芯冷却系统配置研究

基于小型模块化压水堆设计特点，分别采用确定论方法和概率论方法分析了事故工况下的应急堆芯冷却系统配置策略。初步分析表明:确定论方法和概率论方法对于安注箱（ACC）的配置需求存在明显差异；在确保安全目标实现的前提下尽可能简化应急堆芯冷却系统设计，建议可取消安全级的ACC，在非能动堆芯补水箱（CMT）失效后利用其他已有注水水源来平衡设计，如能动余热排出系统增设换料水箱安注功能。      

Power-operated relief valve stuck-open accident and recovery scenarios in the Institute of Nuclear Energy Research integral system test facility

Four scaled small break loss-of-coolant accident (LOCA) tests simulating the pressurizer power-operated relief valves (PORVs) stuck-open accidents and the recovery actions in a pressurized water reactor (PWR) were performed at the Institute of Nuclear Energy Research (INER) integral system test (IIST) facility. The objectives of this study are to verify the effectiveness of emergency operating procedure (EOP) and emergency core cooling system (ECCS) on reactor safety. The break sizes were volumetrically scaled down based on one and all three fully-opened PORVs which is equivalent to 0.23% and 0.69% hot leg flow area of the reference plant. The experimental results indicate that in case of high pressure injection (HPI) system failure, the rapid depressurization of the steam generators is proved to be an effective way in the depressurization of the reactor coolant system and the core cooling. In contrast, if only one HPI charging pump operates normally, which injected half (or minimum) flow rate of normal cooling water, the core cooling can be adequately provided without operating the secondary bleeding during PORV stuck-open transient. This paper also presents the scaling methods for the reduced-height, reduced-pressure (RHRP) IIST facility and the test conditions. The validity of the present scaling methodology is confirmed by the results from previous IIST counterpart tests and comparison of the present results with those of the tests performed at the full-height, full-pressure(FHFP) stuck-open tests.