基于蒙特卡罗方法事故进程分析的CPR1000全厂断电事故PSA

本文对CPR1000核电厂全厂断电事故下可能出现的事故序列的进程建立数学模型，使用蒙特卡罗方法，编写程序计算了每种事故进程中交流电源及时恢复的可能性。根据计算结果对全厂断电事故进行了概率安全分析（PSA）。结果表明，使用蒙特卡罗方法对全厂断电事故进程进行动态分析，可使PSA更贴近核电厂实际情况，有利于更好地认知核电厂整体风险和全厂断电风险。     

辅助给水系统对缓解全厂断电事故能力研究

以CPR1000核电站为研究对象,采用RELAP5/MOD3.2轻水堆瞬态分析程序,对系统进行合理简化并建模,模拟系统在全厂断电事故下的瞬态响应过程,研究全厂断电事故发生后辅助给水（AFW）的投入对缓解全厂断电事故的能力。计算结果表明：断电事故发生后,主给水丧失导致一回路压力和冷却剂平均温度在断电后6s达到峰值;辅助给水投入约200s后,一回路因热阱丧失而引起的温度和压力升高能有效地得到缓解,为交流电源的恢复及余热排出系统的投入赢得了更多的时间。     

CPR1000全厂断电事故瞬态特性分析

用RELAP5/MOD3.4程序对CPR1000压水堆一回路系统进行整体建模,分析全厂断电事故下一回路主要参数的瞬态热工水力特性,并将RELAP5模型计算结果与THEMIS程序的计算结果进行对比,二者符合得较好。计算结果表明：该模型可较准确地模拟CPR1000在事故下的热工水力特性。     

CPR1000全厂断电叠加蒸汽发生器安全阀误开启事故引起的严重事故分析

利用MELCOR程序对CPR1000全厂断电叠加蒸汽发生器(SG)安全阀误开启事故引发的严重事故进行建模与分析,初步实现了对CPR1000严重事故进程的仿真计算与模拟。文中重点分析了无轴封泄漏和辅助给水、有轴封泄漏和辅助给水、有轴封泄漏但无辅助给水3种不同假设条件下CPR1000全厂断电严重事故的响应进程和结果。计算结果显示,SG安全阀误开启对事故进程有重要影响。在无轴封泄漏和辅助给水的情况下,压力容器在9576 s失效;当存在辅助给水时,压力容器失效延后近30000 s;而当存在轴封泄漏时,压力容器失效延后50 s左右。结果证明了发生全场断电叠加SG安全阀误开启事故情况下辅助给水和轴封泄漏对事故起到有效缓解作用。     

非能动核电厂全厂断电事故现象识别与排序研究

为确定整体效应试验模拟中的重要热工水力现象,本文以AP1000为例,对AP系列非能动核电厂全厂断电工况下的事故现象进行了识别和排序。通过分析非能动全厂断电的事故进程划分了事故阶段,并基于核电厂设计进行了系统分解;通过对法规进行技术分析,获得了非能动核电厂全厂断电事故的安全要求和评判指标;通过对主回路冷却剂系统(RCS)、非能动堆芯冷却系统(PXS)内热工水力现象的识别和重要度判断,得到了非能动核电厂全厂断电事故现象识别与排序表。研究结果表明:非能动核电厂全厂断电事故可分为主回路自然循环、非能动堆芯冷却系统自然循环和长期冷却三个阶段;主冷却剂系统的水体积,尤其是稳压器内的水体积是全厂断电事故中应关注的核心评判指标;在系统部件内识别出的热工水力现象,按其对评估指标的影响程度,可进行现象重要度排序。     

CPR1000核电厂全厂断电事故情况下严重事故缓解措施有效性研究

福岛核事故发生后,国内外对严重事故更加重视,严重事故管理导则SAMG的编制和实施已成为监管要求。在建核电厂首次装料前,要制定并实施严重事故管理导则,定期对导则进行修订并验证严重事故管理指南和缓解措施的有效性。本文在调研其他核电机组严重事故缓解措施的基础上,利用严重事故仿真验证系统(VVS),选取全厂断电(Station Blackout,简称SBO)加一回路大破口事故作为CPR1000机组的重要严重事故序列,研究了反应堆功率运行(RP)模式下严重事故缓解措施PSAMG的有效性,重点研究了机组在NS/RRA模式下发生严重事故后,现有导则SSAMG缓解措施的有效性,为CPR1000机组严重事故管理导则SSAMG的完善提供参考。     

大型核电站全厂断电事故的失效模式分析

运用故障树分析方法，对核电站全厂断电事故进行分析。建造了全厂断电事故即厂用电力系统Ａ、Ｂ两列６．６ｋＶ交流应急母线ＬＨＡ和ＬＨＢ同时失电故障树。利用ＳＥＴＳ程序及法国标准９０万千瓦压水堆核电站２００堆年运行经验反馈的可靠性数据，对全厂断电故障树作了定性、定量分析，得到了全厂断电事故的发生概率。给出了全厂断电事故的主要失效模式及发生的概率和事件的重要度。     

CPR1000非能动余热排出系统流动不稳定性分析

利用RELAP5/MOD3.4程序对CPR1000非能动余热排出系统在全厂断电事故(SBO)下的流动特性进行分析,主要分析了非能动余热排出系统空气冷却器的布置方式和空气冷却塔的高度对蒸汽发生器(SG)二次侧流动不稳定性的影响.计算结果表明,水平布置的空气冷却器可以明显减小SG二次侧流动不稳定性；随着空气冷却塔高度增加,...     

非能动核电厂全厂断电事故自然循环现象研究

为识别全厂断电事故下非能动核电厂的主要热工水力现象,对AP1000核电厂全厂断电工况下的事故序列和自然循环现象进行了研究。通过建立AP1000的节点模型,进行了全厂断电事故序列的模拟,并划分了事故阶段,分析了非能动堆芯冷却系统中堆芯补水箱(Core Makeup Tank, CMT)投入失效和安全壳内置换料水箱(In-containment refueling water storage tank, IRWST)参数异常对事故自然循环过程的影响,研究结果表明:全厂断电事故下,非能动核电厂的堆芯衰变热由多个单相自然循环过程导出,其中堆芯与非能动余热排出热交换器(Passive Residual Heat Removal Heat Exchanger, PRHR HX)之间的自然循环对堆芯衰变热的导出具有显著影响。根据热阱的不同和系统参数变化的特点,事故序列可划分为主回路自然循环、非能动堆芯冷却系统(Passive Core Cooling System, PXS)自然循环和长期冷却三个阶段;CMT投入、IRWST水箱参数对PXS自然循环过程存在重要影响。     

M310机组应对全厂断电事故的优化研究

全厂断电事故作为一项超设计基准事故,在核电厂安全分析和设计运行中得到广泛关注。该事故产生的最大风险在于可能丧失堆芯衰变热排出功能,因此如何提高事故期间机组排出堆芯余热的能力,是本事故分析的核心。早在20世纪80年代,美国核管会便发布和实施了联邦法规10CFR 50.63,即全厂断电事故规则及相关技术文件,显著提高了核电厂应对全厂断电事故的能力。本文总结了美国核管会对全厂断电事故的考虑和核电厂的良好实践,对比国内实际,提出国内M310机组应对全厂断电事故的改进建议。     

Thermal-hydraulic design and transient analysis of passive cooling system for CPR1000 spent fuel storage pool

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool(SFP). Our design can effectively manage the SFP temperature not to exceed80 C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout(SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively.The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 C after losing the heat sink.     

Severe accident analysis for a typical PWR using the MELCOR code

Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident.     

PSA of CPR1000 SBO Accident Based on Monte Carlo Accident Process Analysis

A mathematical model for the process of possible accident sequence in the CPR1000 nuclear power plant was established, and the Monte Carlo method was used to programming codes to calculate the possibility of timely recovery of AC power in each accident process. According to the calculation results, probabilistic safety analysis (PSA) of the station black out (SBO) was conducted in the paper. The results show that using Monte Carlo method to analyze the process of SBO can make the PSA more in line with the actual situation of the nuclear power plant. And the overall risk of the nuclear power plant and the risk of SBO can be understood better by using the method.     

Thermal hydraulic characteristics analysis of direct residual heat removal system for China pressurized reactor

An innovative Direct Residual Heat Removal System (DRHRS) is proposed for Pressurized Water Reactor (PWR) in this paper. The new designed parallel DRHRS is different from traditional Passive Residual Heat Removal System (PRHRS), which is connected to steam generation. The thermal hydraulic transient analysis of the new designed DRHRS for CPR1000 has been carried out using the widely accepted safety analysis software RELAP5. The new designed DRHRS is directly connected to the primary loop, which consists of three independent parallel loops, three intermediate cooling circuits and an air loop. The transient behaviors of passive safety system are studied, and design parameter sensitivity analysis is carried out. Results show that during Station Black_Out (SBO) accident, natural circulations are established stably in passive safety system so that core decay is continuously removed from primary loop. And the new designed DRHRS has the capability of removing residual heat to the atmosphere without any external energy input at different surrounding environmental temperature. In emergency, the DRHRS directly remove core decay heat from reactor outlet, and efficiency of residual heat removal is improved. Moreover, reactor power plant maintains safe even if double-ended rupture of a single tube during SBO accident occurs. Thus, the designed DRHRS has great significance for increasing the degree of inherent safety features of CPR1000.     

HPR1000非能动堆腔注水冷却系统事故缓解能力评估

华龙一号（HPR1000）设计了堆腔注水冷却系统（CIS）以实现严重事故期间熔融物的堆内滞留（IVR），该系统分为能动与非能动两列子系统，其中非能动CIS应对的是全厂断电（SBO）始发的严重事故工况。本文对非能动CIS的事故缓解能力进行评估。首先开发了下封头熔池换热计算程序并予以验证，使用MAAP程序对SBO严重事故序列及SBO叠加不同尺寸一回路破口始发的严重事故序列进行计算，并结合熔池换热计算程序得到不同事故序列下的压力容器外壁面最大热流密度，进而评估不同事故序列下非能动CIS的有效性。评估结果表明，非能动CIS可有效应对SBO始发的严重事故序列以及SBO叠加一回路破口尺寸小于60 mm始发的严重事故序列，实现IVR策略。评估结果可应用于HPR1000的严重事故管理。     

非能动余热排出系统敏感性分析

以中国改进型压水堆核电站CPR1000为研究对象,在其蒸汽发生器二次侧设计了一套非能动余热排出系统（PRHRS）,该系统采用在蒸汽发生器二次侧建立自然循环的方式间接带走堆芯余热,确保事故条件下堆芯安全。用RELAP5/MOD3.2程序对系统进行了合理的简化并建模,在全场断电（SBO）事故条件下模拟了PRHRS的瞬态响应过程,并对高位水箱的容积、PRHRS换热器的换热面积、冷热中心高度差以及PRHRS的投入时间等影响PRHRS工作特性的相关参数进行了敏感性分析。计算结果表明：增加高位水箱的容积和增大换热面积均有助于二次侧余热排出系统带走一回路的堆芯余热;降低冷热中心高度差对PRHRS的自然循环能力影响不大;余热排出系统投入时间越早,蒸汽发生器二次侧水位越高,越有利于一次侧余热的排出。