大亚湾核电站安装氢气复合器方案计算分析

针对大亚湾核电站在严重事故条件下安全壳内氢气的产生、释放以及氢气复合器对氢气浓度的影响进行了研究。在假设安全壳内安装一定数量的NIS型被动催化式氢气复合器的条件下,研究了氢气浓度在安全壳内的变化。结果表明,当安全壳内安装大约50个氢气复合器时,最大氢气浓度可以控制在10%左右。     

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。     

核电厂非能动氢复合器研制

核电厂非能动氢复合器主要用于消除严重事故后安全壳内产生的氢气,避免氢气聚集而产生爆炸。根据H2和O2催化反应消氢的工作原理,设计以Pt、Pd混合配比作为催化剂的催化板,并以此为核心部件,设计制造能够在非能动条件下持续消氢的非能动氢复合器。针对核电厂安全壳严重事故后的消氢要求,开展非能动氢复合器在不同温度、压力、氢气体积分数等条件下的消氢速率试验,不同毒物对消氢效果影响试验以及启动和停止阈值试验。研究结果表明,非能动氢复合器达到了核电厂事故后消氢技术要求,可直接应用于二代堆型核电厂,还可以应用于EPR或AP1000等三代堆型核电厂。     

氢气缓解措施中点火器特点及有效性分析

为保证严重事故下安全壳的完整性，氢气缓解措施广泛应用于核电站内。本文应用三维计算流体力学程序GASFLOW分析了氢气缓解措施中的点火器系统与复合器系统，并总结出各自的特点。点火器通过点燃的方式能够快速有效地降低氢气总量，同时会明显增大安全壳内压力与温度；复合器需长时间运行才能够消除大量的氢气，工作的同时不会引起平均温度与压力的明显上升。如果点火器的布置位置及启动时间均合理，有可能在不引起大范围火焰加速或爆炸的情况下迅速有效地消除氢气。     

严重事故管理导则中的氢气风险判断及事故模拟分析

建立严重事故管理导则中用于判断氢气燃烧、超压风险以及安全壳降压时氢气风险的判断工具。用一体化事故分析程序对全厂断电事故进行模拟计算,用该氢气风险判断工具对不同事故阶段的氢气风险进行分析。结果表明:在全厂断电始发的严重事故下,没有氢气复合器且没有安全壳喷淋时,安全壳大气在一段时间内会被水蒸气惰化,不会发生燃烧,但如果应急电源恢复,重新启动安全壳喷淋时,有可能引起氢气燃烧甚至造成安全壳超压;在增加氢气复合器后,没有造成安全壳超压的风险,并且判断结果是保守的。     

基于CFD的安全壳局部隔间非能动氢气复合器布置方案研究

非能动氢气复合器用于压水堆核电厂严重事故条件下安全壳内氢气的消除。通过计算流体力学(CFD)方法能够给出事故条件下非能动氢气复合器周围三维流场和温度场的分布。基于CFD程序根据非能动氢气复合器消氢公式,计算非能动氢气复合器进出口的气体流量和气体组分,并作为非能动氢气复合器的边界条件,开展三维空间内非能动氢气复合器消氢速率和氢气分布情况研究。结果表明：简化的非能动氢气复合器模拟方案能很好地模拟非能动氢气复合器样机的消氢效果;对安全壳内局部隔间开展非能动氢气复合器消氢效果研究发现,在相同环境条件下,非能动氢气复合器布置在较高位置与布置在较低位置相比,布置在较高位置时,非能动氢气复合器具有更高的消氢速率,隔间整体氢气浓度较低,但是非能动氢气复合器布置在较高位置时出现隔间底部局部氢气聚集的情况。     

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.     

900MW核电站严重事故缓解系统功能分析

应用新版的MELCOR程序，以900MW机组为对象，进行了SBO严重事故进程研究，在严重事故计算分析中比较了稳压器功能延伸，非能动氢气复合等缓解措施（3个方案）对严重事故进程和现象的影响。对堆芯熔融过程中包壳和燃料栅元的径向和轴向分段失效模式进行了模拟；计算了MCCI引起的堆坑径向和轴向熔蚀的情况；对事故中后期可燃气体的产生、分布及非能动氢气复合系统在安全壳中对氧气的复合效应进行了评价和分析。分析结果表明，事故下稳压器延伸功能的及时投入，可使堆芯整体坍塌失效及压力容器熔穿均延后了2h左右，并且避免了高压堆熔引起的安全壳直接加热现象，消除了由此引起的对安全壳完整性的威胁。各方案均表明，由于从一回路迁移到安全壳的大量水蒸汽对氢气燃烧的惰化作用，在一定程度上避免了安全壳内氢爆的发生，而氢气复合器在空间和数量上的合理布置，则可以完全消除大空间内燃爆的威胁。24h内堆坑地板没有完全熔穿的情况出现。     

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。     

大亚湾和岭澳核电站安全壳氢气风险及消氢措施有效性分析

使用MAAP程序计算大亚湾和岭澳核电站严重事故条件下安全壳内的相关质能释放和氢气源项;利用TONUS程序建立安全壳集总参数模型,计算分析氢气在安全壳内的分布情况;结合非能动氢复合器消氢性能、现场条件和氢气分布情况,提出氢复合器布置方案;借助TONUS和GASFLOW程序,分别使用集总参数法和CFD法,验证消氢方案的有效性。验证结果表明,安全壳内氢气浓度满足相关法规要求。     

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。     

On the failure modes of alternative containment designs following postulated core meltdown

The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR representative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best estimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. These include the temperature of the non-condensibles escaping the ice condenser into the upper compartment, the performance of the pressure suppression system, the distribution of non-condensibles between compartments, and the degree and rate of combustion of hydrogen generated from metal-water reactions. For the PWR ice condenser case, results indicate that the containment would be breached by (i) steam overpressurization during the blowdown period (time less than 20 sec) if the ice condenser fails to perform its function, (ii) by overpressurization and thermal stress during the core melt period if 25% or more of the core zirconium reacts with water followed by hydrogen burning and, and (iii) by the overpressurization due to non-condensibles before containment floor penetration is completed. For the BWR Mark III case, similar conclusions can be drawn for the loss of vapor suppression, and for the hydrogen burning if the extent of zirconium-water reaction is more than 35% of the core inventory. If the hydrogen burning fails to materialize, the containment can retain its integrity until containment meltthrough provided the melting is confined to the reactor pedestal area. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded.     

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.     

Hydrogen and steam distribution following a small-break LOCA in large dry containment

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations. Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident （LOCA） without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump （RCP） compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.     

蒸汽份额对安全壳内氢气分布的影响

核电厂严重事故下安全壳内氢气的热工水力特性极其复杂,安全壳内氢气的流动与分布受多种因素影响,如安全壳通路、产氢速率、水蒸气份额等。本文使用三维计算流体力学软件CFX研究安全壳内的氢气浓度分布,关注在产生的混合气体中水蒸气份额对安全壳内氢气分布的影响。研究结果表明：所产生的混合气体中的水蒸气份额越高,水蒸气从破口区域携带出来的氢气越多;水蒸气促进了安全壳内的空气流动,导致破口区域的氢气浓度较低,其他区域的氢气分布则较为均匀。     

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.     

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.     

大型干式安全壳消氢系统的初步设计

以岭澳核电站为分析对象,利用MELCOR和TONUS(CEA)程序进行分析计算,给出了初步的消氢系统设计方案,对不同核电站的消氢系统设计方案进行了对比和讨论.结果表明:安全壳内安装33个FR750型或者17个左右的FR1500型氢气复合器可以满足氢气控制要求.     

Direct containment heating integral effects tests in geometries of European nuclear power plants

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P’4, the VVER-1000 and the German Konvoi plant. A high-temperature iron–alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified.The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.