核电厂安全壳地坑滤网堵塞问题的机理分析

破口失水事故工况下,大量碎片可能随着泄漏的冷却剂和喷淋液迁移到安全壳地坑滤网处,并逐渐堆积形成碎片床,不断增大流体通过滤网的阻力,降低应急堆芯冷却系统或安全壳喷淋系统泵的净正吸入压头裕量并导致堆芯、安全壳丧失冷却,从而威胁核电厂安全。本文对核电厂发生假想破口失水事故后碎片的产生、迁移,以及在安全壳地坑滤网处堆积成碎片床,并造成地坑滤网堵塞的机理进行分析说明。     

核电厂安全壳地坑滤网堵塞问题的纠正措施研究

核电厂发生破口失水事故后,当应急堆芯冷却系统或安全壳喷淋系统处于再循环模式运行时,碎片堵塞对安全壳地坑滤网的性能存在着潜在的影响,而且碎片迁移过程中的堵塞可能会对再循环模式需要的流道造成不利的影响。本文将从碎片产生、碎片输运和地坑滤网设计等方面,论述说明针对地坑滤网堵塞问题可能采取的纠正措施。     

压水堆安全壳地坑滤网化学效应WCAP分析方法

针对核电厂地坑滤网安全性能问题,美国核管理委员会(NRC)先后出台了一系列RG1.82"失水事故后长期再循环冷却的水源"管理导则的修订版,用以指导地坑滤网堵塞研究。冷却剂失水事故(LOCA事故)后在安全壳喷淋液和安全壳地坑介质的化学环境会导致安全壳内的各种碎渣中化学元素的溶解,并且随着安全壳地坑介质温度的降低形成沉淀析出,所析出的沉淀会在安全壳地坑滤网表面物理碎渣床上形成二次沉积,从而造成滤网压损性能的进一步恶化,此即为安全壳地坑滤网的化学效应。本文介绍压水堆安全壳地坑滤网化学效应的试验分析方法。     

压水堆核电厂安全壳内碎片源的踏勘方法介绍

安全壳地坑是许多压水堆核电厂设计为在失水事故后为堆芯冷却和安全壳排热提供再循环水的专设安全设施。安全壳内的潜在碎片源在事故中可能堵塞安全壳内的地坑滤网,从而造成安全壳地坑性能下降。为了评价安全壳地坑在破口事故后能否满足设计要求,首先应确定潜在碎片源的类型以及它们在安全壳内的位置。安全壳内现场踏勘就是寻找与定位碎片源的有效方法,并能够提供一些进行安全壳地坑性能分析的必要信息。介绍了压水堆核电厂安全壳内碎片源的一些踏勘方法。     

先进压水堆核电厂安全壳地坑滤网设计

核电厂的地坑滤网堵塞问题是目前世界核能领域比较关注的重要安全问题之一,先进压水堆的设计也充分体现了改善地坑性能的最新研究成果。本文介绍了在我国已经开工建设的先进压水堆核电厂（AP1000以及EPR）的地坑滤网设计,可以为我们解决地坑滤网堵塞问题提供有益的借鉴。     

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。     

核电厂安全壳内置换料水箱过滤系统过滤性能及阻力特性研究

对核电厂安全壳内置换料水箱（IRWST）过滤系统过滤性能及压损特性进行了分析研究。该研究借助试验验证和数值模拟分析方式展开,主要包括下游效应（碎片浓度）试验研究和过滤系统压损分析2部分。结果表明,在事故工况下,地坑滤网下游碎片浓度为368 ppm（1 ppm=1 mg/L）、安全注入系统（RIS）地坑滤网和安全壳喷淋系统（EHR）地坑滤网的压损分别为3.533 kPa和3.631 kPa,上述结果分别满足了过滤系统碎片浓度小于480 ppm和压损小于5.6 kPa的系统功能要求。      

安全壳地坑过滤器下游效应(堆芯外)分析

在安全壳地坑过滤器纤维层形成的过程中,颗粒和部分纤维材料穿过过滤器进入下游的安全注入系统(RIS)/安全壳喷淋系统(EAS)流道,核安全监管单位要求对此展开评估,确定碎渣对于过滤器下游设备的影响。以秦山核电二期扩建工程为例,使用评估文件WCAP-16406中推荐的通用方法,对压水堆核电厂中地坑过滤器下游(堆芯外)设备的堵塞和磨损影响进行相应的计算和分析。     

核电厂安全壳地坑过滤器化学效应试验研究

核电厂失水事故工况下,化学效应可能引起地坑过滤器过度堵塞,影响应急堆芯冷却系统或安全壳喷淋系统的再循环功能,导致堆芯、安全壳丧失冷却,威胁核电厂的安全。本文以秦山核电厂二期扩建工程为研究对象,开展失水事故工况下潜在化学产物对地坑过滤器压头损失的影响研究。结果表明,秦山核电厂二期扩建工程安全壳内含Al材料和保温材料在地坑环境中会析出Al、Si,Al、Si元素在地坑介质降温过程中形成了化学沉淀物,化学沉淀物会在地坑过滤器碎片床上沉积,堵塞碎片床流道,缩小碎片床孔隙率,导致地坑过滤器压头损失增加。因此,秦山核电厂二期扩建工程失水事故后存在化学效应,在地坑过滤器性能评估、下游效应分析中应予以考虑。      

CPR1000核电厂安全壳地坑滤网化学效应研究

研究了在冷却剂丧失事故(LOCA)工况下CPR1000核电厂安全壳地坑滤网堵塞的化学效应问题。化学效应评价采用模拟试验的方法进行。结果表明,在CPR1000核电厂中化学效应对应急再循环的影响非常轻微。     

Material corrosion in a reactor containment sump following a loss-of-coolant accident

The present study conducts a critical review of relevant research that pertains to Generic Safety Issue 191 (GSI-191) and the chemistry and corrosion behavior of the various materials present in the containment sump of a pressurized water reactor (PWR). Thermodynamic simulations, integrated tests, and benchtop tests have determined the structure of the potential precipitates that may clog the sump strainer and cause a failure of the emergency core cooling system (ECCS). Based on this research, it was determined that the available research has been insufficient in terms of simulating the transient temperature behavior in the containment sump in the post loss-of-coolant environment. Research gaps are identified, and recommendations for future research are presented.     

Evaluation of recirculation sump performance for OPR1000 plant: Part I debris transport during the blow-down phase of LOCA

After TMI-2 accident, long-term core cooling management takes more importance rather than short-term management since probabilistic safety assessment performed revealed that long-term management had higher risk than the risk from short-term management. Regarding to this, since 1992, blockage of sump suction strainer was taken a place in Barseback Unit 2 (Sweden, BWR) and the United States Nuclear Regulatory Commission (USNRC) took into consideration as a General Safety Issue 191 (GSI-191), “PWR Sump Blockage,” in 1998 and opening an investigation.As a response to GSI-191 the Nuclear Energy Institute (NEI) submitted a recirculation sump evaluation methodology called NEI 04–07. In this methodology 0.75 was recommended for blow-down transport to lower containment based on the study on a boiling water reactor sump clogging issue and engineering judgment. USNRC quantitatively evaluated the blow-down transport in the safety evaluation report to NEI 04–07, and concluded the recommended blow-down transport fraction in NEI 04–07 was sufficiently conservative. However, the safety evaluation report to NEI 04–07 includes many values in evaluation steps which depend much on the containment configuration and engineering judgment. In particular the dependency on the plant type limits the generality of the USNRC’s conclusion on blow-down transport, when the considered plant is different from the volunteer plant.This study provides a modified evaluation method and results of debris blow-down transport for an Optimized Power Reactor 1000 MWe (OPR1000) in Korea. The modified method includes more realistic physical background with less uncertainty while providing consistent result with NEI and USNRC methods.     

严重事故下安全壳内环境条件计算分析

参照对先进压水堆安全壳的要求,结合恰希玛二期工程严重事故缓解措施,对大破口失水事故(LLOCA)叠加安注失效、小破口失水事故(SLOCA)叠加安注失效、全厂断电(SBO)叠加柴油机驱动的辅助给水失效等严重事故序列可能影响安全壳内环境的条件及缓解措施进行了分析.结果表明,恢复喷淋可以明显地降低安全壳内的压力和温度,有效地改善安全壳内的环境,从而改善各种仪表设备的工作条件.     

循环水浸泡影响下的新型地坑过滤器抗震分析

安全壳地坑过滤器是核电厂专用的重要安全设施,用于在失水事故（LOCA）下收集安全壳内的泄漏水和喷淋水,以便再循环使用,从而保证安注泵和喷淋系统不被杂质损坏。为了对新型地坑过滤器结构的抗震性能进行分析,以其中的典型模块为例,采用ANSYS软件进行数值建模,并针对其在不同工况下进行了应力分析;地坑过滤器结构的附加水质量通过经验公式和试验的方法得到,在抗震分析中考虑了地坑过滤器结构周围流体对结构抗震性能的影响;并依据RCC-M规范对过滤器的滤筒和汇流槽进行了应力强度评定。该模块的计算结果表明,在不同工况下该新型安全壳地坑过滤器具有良好的抗震性能,满足规范的对强度的要求。     

Effect of spray on performance of the hydrogen mitigation system during LB-LOCA for CPR1000 NPP

During the course of the hypothetical large break loss-of-coolant accident (LB-LOCA) in a nuclear power plant (NPP), hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel (RPV). It is then ejected from the break into the containment along with a large amount of steam. Management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen mitigation system (HMS) and spray system in CPR1000 NPP. The computational fluid dynamics (CFD) code GASFLOW is utilized in this study to analyze the spray effect on the performance of HMS during LB-LOCA. Results show that as a kind of HMS, deliberate igniter system (DIS) could initiate hydrogen combustion immediately after the flammability limit of the gas mixture has been reached. However, it will increase the temperature and pressure drastically. Operating the DIS under spray condition could result in hydrogen combustion being suppressed by suspended droplets inside the containment. Furthermore, the droplets could also mitigate local the temperature rise. Operation of a PAR system, another kind of HMS, consumes hydrogen steadily with a lower recombination rate which is not affected noticeably by the spray system. Numerical results indicate that the dual concept, namely the integrated application of DIS and PAR systems, is a constructive improvement for hydrogen safety under spray condition during LB-LOCA.