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.     

百万千瓦级压水堆严重事故下氢气源项及氢气控制有效性分析

针对百万千瓦级压水堆核电厂大型干式安全壳在严重事故情况下的氢气风险控制,建立了一体化事故分析模型,分别对大破口失水事故(LB-LOCA)、中破口失水事故(MB-LOCA)、小破口失水事故(SB-LOCA)、全厂断电事故(SBO)、蒸汽发生器(SG)传热管破裂事故(SGTR)以及主蒸汽管道破裂事故(MSLB)进行事故进程计算以及氢气源项分析。相对于其他事故序列,LB-LOCA下堆芯快速熔化,锆-水反应产生氢气的速率快,可以作为安全壳内氢气风险控制有效性分析的代表性事故序列。分析表明,严重事故情况下在安全壳中安装一定数量的非能动氢气复合器(PARs)能够有效去除安全壳中的氢气,消除氢气燃烧或爆炸的风险,保持安全壳的完整性。     

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。     

典型严重事故非能动安全壳冷却系统效果分析

先进压水堆采用非能动安全壳冷却系统(PCCS)在事故下维持安全壳完整性,包括重力喷洒形成安全壳外部水膜冷却和空气冷却流道中空气对流传热。针对严重事故下PCCS效果研究,建立了非能动压水堆安全壳及非能动安全壳冷却系统的传热分析模型(包括对流传热及蒸发/冷凝传热),并耦合反应堆主系统模型及专设安全设施模型。通过与西屋公司PCCS大尺度试验结果的比对验证了模型的可用性,进而针对非能动先进压水堆选取全厂断电、热段小破口失水始发事故作为典型严重事故序列,模拟了事故进程、主系统响应及安全壳的响应,分析了PCCS对安全壳的降温、降压作用。结果表明,安全壳压力72h内未超过安全限值,保持安全壳完整性。     

压水堆核电厂消氢装置选用探讨

为消除核电厂在严重事故工况下积聚在安全壳内的氢气,需要在核电厂安全壳内增加消氢装置。本文分析了严重事故工况下目前主要采用的氢气点火器和非能动氢复合器的原理,并结合某核电厂增加非能动氢复合器改造的工程实践,给出了压水堆核电厂消氢装置选用方案。实践结果表明,此最优消氢方案既可以保证电厂安全,又可以节约成本,具有巨大的社会效益和经济效益。     

秦山二期核电厂严重事故下安全壳内氢气浓度分布及风险初步分析

采用模块化严重事故计算工具,对秦山二期核电厂大破口失水事故(LB-LOCA)、小破口失水事故(LB-LOCA)和全厂断电(SBO)诱发的严重事故序列以及安全壳内的氢气浓度分布进行了计算分析.在此基础之上,参考美国联邦法规10CFR关于氢气控制和风险分析的标准,对安全壳的氢气燃烧风险进行了初步研究.分析结果表明:大破口严重事故导致的安全壳内的平均氢气浓度接近10%,具有一定的整体性氢气燃烧风险,小破口失水和全厂断电严重事故可能不会导致此类风险,但仍然存在局部氢气燃烧的可能.     

国产先进压水堆严重事故下氢气行为及控制系统分析

严重事故下的氢气控制是核电厂安全需要考虑的重要问题之一。采用一体化严重事故分析程序对国产先进压水堆核电厂进行系统建模,选取大破口触发的严重事故序列,对严重事故工况下的氢气产生情况及氢气控制系统的性能进行分析评价。结果表明:大破口事故序列下氢气的产生主要有两个阶段,分别是早期锆包壳与水反应产生氢气及堆芯熔融物迁移至下腔室产生氢气,其中燃料包壳的氧化是产氢的主要阶段,氢气释放时间较早,氢气产生速率较大。氢气控制系统的设计能够有效缓解可能的氢气风险,满足相关法规标准的安全要求,确保安全壳的完整性。     

重水堆核电厂典型严重事故氢气风险分析

核电厂严重事故下的氢气控制一直是核电厂关注的热点问题之一。本文采用重水堆一体化事故分析程序建立了主热传输系统(PHTS)模型、排管容器及端屏蔽系统、堆腔以及安全壳模型。分别选取代表高压熔堆和低压熔堆的全厂断电及出口集管大破口失水事故始发严重事故序列,从堆芯氧化产氢以及系统热工水力行为出发,对重水堆产氢特性及点火器的消氢效果进行了研究。分析表明:严重事故下随着堆芯冷却恶化,排管容器内发生锆水反应而产生氢气,排管容器和堆腔内的水对氢气产生有较长时间的抑止作用,随着排管容器和堆腔内水的逐渐烧干,排管容器蠕变失效,熔融堆芯落入堆腔发生堆芯熔融物与混凝土的相互作用而产生大量氢气。当氢气点火器失效时,安全壳隔间内氢气体积份额持续增加,存在燃爆风险;点火器开启时,隔间中的氢气混合气体在较低浓度下点燃,氢气燃烧模式处于慢速燃烧区。     

基于Gasflow程序的非能动安全压水堆氢气行为计算和分析

本文利用Gasflow程序对非能动压水堆发生假想的严重事故后,安全壳内的氢气流动、分布和积聚行为进行了计算和分析,对安全壳内各房间的氢气风险进行了评价并给出了降低氢气燃烧风险的建议。计算结果表明,在发生大破口事故中,安全壳内氢气浓度较高的区域为破损蒸汽发生器隔间,内置换料水箱隔间和上部隔间,需要设置消氢系统来降低隔间内的氢气浓度。     

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.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

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.     

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。     

Severe accident analysis of a 1300 MWe PWR to address the hydrogen issue

The behaviour of the potentially large quantity of hydrogen generated during a severe accident has been recognised as an issue of importance since the accident at Three Mile Island. In this article, we describe a severe accident analysis for the Neckarwestheim 2 1300 MWe PWR “Konvoi” plant, performed primarily to investigate the behaviour of hydrogen in the containment, and draw conclusions regarding the need for hydrogen control systems (igniters). The Modular Accident Analysis Program (MAAP) developed by IDCOR in the United States, and the Westinghouse COMPACT multi-compartment containment code were used. The study investigated the generation, release to containment, distribution within containment and potential combustion of hydrogen produced during two severe accident sequences. Results are summarized which show that hydrogen mixing in containment is generally good and that even without hydrogen control systems, hydrogen combustion, although possible, does not threaten containment integrity.     

AP1000安全壳外环廊氢气风险分析

本文采用MAAP程序对AP1000核电厂的环廊区域进行建模,计算严重事故下的氢气浓度,以合理评估壳外氢气爆炸风险。分析结果表明：AP1000核电厂所设置的氢气点火器和氢气复合器能很好地控制环廊氢气浓度,防止壳外氢气风险的发生。只有在氢气点火器和氢气复合器均不可用,且产氢量很大的极限工况下,才可能在环廊区域内出现较高的氢气浓度,威胁安全壳的完整性。     

严重事故氢气燃爆缓解措施的初步研究

轻水堆核电站发生严重事故时,氢气的大体积氢燃爆可能会严重威胁安全壳的完整性.氢气点火器与氢气复合器是2种严重事故下的氢气燃爆缓解设备.本文分别研究了3种氢气燃爆缓解措施,包括仅采用氢气点火器、仅采用氢气复合器和采用氢气复合器结合点火器.结果表明,采用氢气复合器结合点火器的方式可以安全、持续、有效地降低大体积氢燃爆带来的风险.     

基于氢气风险分析的ADS手动卸压策略研究

先进非能动压水堆设计采用自动卸压系统(ADS)对一回路进行卸压,严重事故下主控室可手动开启ADS,缓解高压熔堆风险。然而ADS的设计特点可能导致氢气在局部隔间积聚,带来局部氢气风险。本文基于氢气负面效应考虑,对利用ADS进行一回路卸压的策略进行研究,为严重事故管理提供技术支持。选取全厂断电始发的典型高压熔堆严重事故序列,利用一体化事故分析程序,评估手动开启第1~4级ADS、手动开启第1~3级ADS、手动开启第4级ADS 3种方案的卸压效果,并分析一回路卸压对安全壳局部隔间的氢气负面影响。研究结果表明,3种卸压方案均能有效降低一回路压力。但在氢气点火器不可用时,开启第1~3级ADS以及开启第1~4级ADS卸压会引起内置换料水箱隔间氢气浓度迅速增加,可能导致局部氢气燃爆。因此,基于氢气风险考虑,建议在实施严重事故管理导则一回路卸压策略时优先考虑采用第4级ADS进行一回路卸压。     

大功率非能动压水堆严重事故工况堆芯熔毁进程研究

应用MELCOR 2.1程序,建立了大功率非能动反应堆主要回路、非能动安全系统及安全壳的热工水力模型,并以热段小破口叠加ADS 1阀门失效和内置换料水箱失效触发严重事故为研究对象,对事故进程进行模拟,对堆芯熔毁进程进行了分析。分析结果表明:1)锆合金和不锈钢氧化释热功率在蒸汽充足的情况下高于燃料的衰变功率,将加速堆芯的恶化;2)约13.1%的不锈钢和27.1%的锆合金被氧化,共产生550.99kg氢气;3)堆芯构件的熔化主要依赖于材料自身的熔点和有无构件支撑,堆芯支撑板能够延缓熔融物跌落进入下封头的进程;4)熔池形成后若外部冷却的不足将很快导致下封头应力失效。     

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.