Analysis of hydrogen risk mitigation with passive autocatalytic recombiner system in CPR1000 NPP during a hypothetical station blackout

Hydrogen safety has attracted extensive concern in severe accident analysis especially after the Fukushima accident. In this study, a similar station blackout as happened in Fukushima accident is simulated for CPR1000 nuclear power plant (NPP) model, with the computational fluid dynamic code GASFLOW. The hydrogen risk is analyzed with the assessment of efficiency of passive autocatalytic recombiner (PAR) system. The numerical results show that the CPR1000 containment may be damaged by global flame acceleration (FA) and local detonation caused by hydrogen combustion if no hydrogen mitigation system (HMS) is applied. A new condensation model is developed and validated in this study for the consideration of natural circulation flow pattern and presence of non-condensable gases. The new condensation model is more conservative in hydrogen risk evaluation than the current model in some compartments, giving earlier starting time of deflagration to detonation transition (DDT). The results also indicate that the PAR system installed in CPR1000 could prevent the occurrence of the FA and DDT. Therefore, HMS such as PAR system is suggested to be applied in NPPs to avoid the radioactive leak caused by containment failure.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

秦山二期核电站氢气风险的CFD研究

利用计算流体力学(CFD)程序GASFLOW模拟了波动管大破口事故发生后7 000 s内装有22台氢气复合器的秦山二期核电站安全壳内的水蒸汽及氢气行为,得到了不同阶段的特征性流场及氢气浓度的分层情况,给出了所采用的复合器布置方案的稳定消氢速率为20 g/s,并指出了破口所在蒸汽发生器隔间内发生氢气燃烧火焰加速的可能性.同时,计算结果表明,安全壳内构筑物吸热带走了大部分从一回路释放的热量;压力变化同时受气体总质量(主要是水蒸汽质量)与温度的控制.     

先进压水堆大破口始发严重事故下安全壳内氢气风险分析

本文采用集总参数法,在先进非能动压水堆核电厂严重事故一体化分析模型基础上,考虑先进压水堆非能动安全特性以及严重事故下采取熔融物堆内滞留(IVR)措施等特性对氢气风险的影响,开展了典型严重事故下安全壳内氢气风险分析。分别选取了冷段双端剪切断裂大破口、冷段大破口叠加IRWST重力注水有效以及ADS-4误启动三个典型大破口失水事故序列,对事故进程中的氧化温度、产氢速率以及产氢质量等特性进行了研究。选取产氢量最大的冷段大破口叠加IRWST重力注水有效事故序列,分析了氢气点火器系统的消氢效果。结果表明,堆芯再淹没过程产生大量氢气,采用点火器可有效去除安全壳内的氢气,从而降低氢气燃爆风险。     

A study on evaluating a passive autocatalytic recombiner PAR-system in the PWR large-dry containment

A systematic step-by-step framework for analyzing hydrogen behavior and implementing passive autocatalytic recombiners (PARs) to mitigate hydrogen deflagration or detonation risk in severe accidents (SAs) is presented. The procedure can be subdivided into five main steps: (1) modeling the containment based on the plant design characteristics, (2) selecting the typical severe accident sequences, (3) calculating the hydrogen generation including in- and ex-vessel period, (4) modeling the gas distribution in containment atmosphere and estimating the hydrogen combustion modes and (5) evaluating the efficiency of the PAR-system to mitigate the hydrogen risk with and without catalytic recombiners, according to the safety criterion. For the Chinese 600MWe pressurized water reactor (PWR) with a large-dry containment, large break loss-of-coolant accident (LB-LOCA) is screened out as the reference severe accident sequence, considering the nature of hydrogen generation and the probabilistic safety assessment (PSA) result on accident sequences. The results show that a certain number of recombiners could remove effectively hydrogen and oxygen, to protect the containment integrity against hydrogen deflagration or detonation.     

先进压水堆核电厂氢气控制策略分析研究

新建核电厂的设计必须做到“实际消除”早期与大量放射性释放的可能性,氢气燃爆导致的安全壳失效是必须要“实际消除”的严重事故工况之一。因此对各种消氢措施的特点进行分析研究,建立联合消氢策略评价方法,可为先进压水堆核电厂氢气控制策略选择设计评价提供支持手段。根据严重事故管理中对氢气控制策略的考虑,研究安全壳内局部位置的可燃性是相关设计评价的关键问题。根据可燃性准则、火焰加速准则、燃爆转变准则,本文使用三维CFD程序对典型严重事故工况下安全壳蒸汽发生器隔间内的可燃性及氢气风险进行模拟分析。研究结果表明,虽然喷放源项中有大量水蒸气,蒸汽发生器隔间中仍有较大区域处于可燃限值以内,合理布置的点火器能在设计中点燃并消除氢气。本研究建立的分析方法能用于对核电厂氢气控制策略选择设计的评价。     

核电厂安全壳地坑滤网堵塞问题研究进展与现状

安全壳地坑是压水堆核电厂专设安全设施安全壳喷淋系统和安全注入系统的重要组成部分。失水事故后安全壳地坑滤网的堵塞将极大地影响安全壳喷淋系统和安注系统的正常运行,威胁核电厂的安全。许多核电国家针对地坑滤网堵塞问题进行了研究,各国核安全管理当局也发布了公告要求核电厂解决此问题。本文介绍了安全壳地坑滤网堵塞问题的研究进展及现状。     

Analysis on containment depressurization under severe accidents for a Chinese 1000 MWe NPP

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment.     

Conceptual design and analysis of a semi-passive containment cooling system for a large concrete containment

An internal evaporator-only (IEO) concept has been developed as a semi-passive containment cooling system for a large dry concrete containment. The function of this system is to keep the containment integrity by maintaining the internal pressure not to exceed ultimate design pressure, i.e. 0.83 MPa (120 psia) in the absence of any other containment cooling following a severe accident, which postulates core damage and hydrogen combustion. The ability of the concept to protect the containment was evaluated for the design basis accident (DBA) large break loss of coolant accident (LB LOCA) and severe accident scenarios (LB LOCA without Emergency Core Cooling System (ECCS) and containment spray flow, 100% zirconium oxidation and complete hydrogen combustion). All were modeled using the GOTHIC computer code. It was concluded that a practical system requiring four IEO loops could be utilized to meet design criteria for severe accident scenarios.     

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.     

先进压水堆核电站氢气风险分析

核电厂在严重事故期间会产生大量氢气并释放到安全壳内,威胁安全壳的完整性。应用氢气风险分析程序GASFLOW对先进压水堆核电站在大破口失水事故叠加应急堆芯冷却系统失效导致的严重事故期间的氢气行为及风险进行分析。结果表明,当气体释放源位于蒸汽发生器隔间时,氢气流动的主要路径为"蒸汽发生器隔间—穹顶空间—操作平台以下隔间";破口隔间的氢气体积浓度分布与源项氢气体积浓度及射流形态有关,非破口区域的氢气体积浓度呈层状分布,在扩散作用下,层状分布向下推移;蒸汽发生器隔间存在着火焰加速(FA)的可能性,但基本可排除燃爆转变(DDT)的可能性,穹顶区域基本可排除FA和DDT的可能性。     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

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.     

Improvement of HYCA3D code and experimental verification in rectangular geometry

Concerns about the local hydrogen behavior in a nuclear power plant (NPP) containment during severe accidents have increased with the 10CFR50.34(f) regulation after TMI accident. Consequently, investigations on the local hydrogen behaviors under severe accident conditions were required. An analytical model named HYCA3D was developed at Seoul National University (SNU) to predict the thermodynamics and the three dimensional behavior of a hydrogen/steam mixture, within a subdivided containment volume following hydrogen generation during a severe accident in NPPs. In this study, the HYCA3D code was improved with a steam condensation and spray model, and verified with hydrogen mixing experiments executed in a SNU rectangular mixing facility. Helium was used to simulate hydrogen in both the calculations and the experiments. The calculation results show good agreement with the experimental data.     

Analysis of hydrogen flame acceleration in APR1400 containment by coupling hydrogen distribution and combustion analysis codes

This study was conducted as part of the construction of an integrated system to mechanistically evaluate flame acceleration characteristics in a containment of a nuclear power plant during a severe accident. In the integrated analysis system, multi-dimensional hydrogen distribution and combustion analysis codes are used to consider three-dimensional effects of the hydrogen behaviors. GASFLOW is used for the analysis of a hydrogen distribution in the containment. For the analysis of a hydrogen combustion in the containment, an open-source CFD (computational fluid dynamics) code OpenFOAM is chosen. Data of the hydrogen and steam distributions obtained from a GASFLOW analysis are transferred to the OpenFOAM combustion solver by a conversion and interpolation process between the solvers. The combustion solver imports the transferred data and initializes the containment atmosphere as an initial condition of a hydrogen combustion analysis. The turbulent combustion model used in this study was validated by evaluating the F22 test of the FLAME experiment. The coupled analysis method was applied for the analysis of a hydrogen combustion during a station blackout accident in an APR1400. In addition, the characteristics of the flame acceleration depending on a hydrogen release location are comparatively evaluated.     

核电站安全壳内氢气扩散和燃烧的分析程序GASFLOW及其应用

介绍了由美国洛斯阿拉莫斯实验室(LANL)和德国卡尔斯鲁厄研究中心(FzK)共同开发的三维计算流体力学程序GASFLOW的基本数学物理模型和数值计算方法。该程序主要用于分析核电站严重事故下安全壳内氢气、水蒸气扩散分布和燃烧。列举了该程序在德国Konvio型压水堆氢气安全分析中的应用。     

严重事故下喷淋模式的研究

核电厂发生严重事故后,安全壳内有可能堆积大量的氢气,如果此时不适宜地投入喷淋,会破坏安全壳内的惰性环境而引起氢气燃烧或者爆炸,甚至导致安全壳失效。为避免氢燃,研究者通过合理的假设,根据相关的实验公式,推导出不同氢气产量下安全壳内压力所需满足的条件,获得了根据安全壳内压力值来指导喷淋开闭的保守的控制模式。文章以大亚湾核电站为分析对象,利用MELCOR来进行分析,验证了此控制模式的可行性,并讨论了堆腔注水、氢气自燃以及安全壳底板成分对制订喷淋模式的影响。     

严重事故下氢气风险及氢气控制系统的初步分析

采用一体化严重事故分析工具,对600MWe压水堆核电厂严重事故下氢气风险及拟定的氢气控制系统进行分析。结果表明：相对于小破口失水始发事故和全厂断电始发事故工况,大破口失水始发严重事故堆芯快速熔化,在考虑100%锆 水反应产氢量的条件下,大破口失水始发事故氢气风险较大,有可能发生氢气快速燃烧;在氢气控制系统作用下,发生大破口失水始发严重事故时,安全壳内平均氢气浓度和隔间内氢气浓度低于10%,未达到氢气快速燃烧和爆炸的条件,满足美国联邦法规10CFR中关于氢气控制和风险分析的准则,认为该氢气控制系统是可行、有效的。     

严重事故下喷淋模式的研究

核电厂发生严重事故后．安全壳内有可能堆积大量的氢气．如果此时不适宜地投入喷淋．会破坏安全壳内的惰性环境而引起氢气燃烧或者爆炸．甚至导致安全壳失效。为避免氢燃．本文通过合理的假设．根据相关的实验公式推导出不同氢气产量下安全壳内压力所需满足的条件,获得了根据安全壳内压力值来指导喷淋开闭的保守的控制模式。本文以大亚湾核电站为分析对象,利用MELCOR来进行分析．验证了此控制模式的可行性．并讨论了堆腔注水、氢气自燃以及安全壳底板成分对制订喷淋模式的影响。     

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。