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。     

乏燃料水池氢气风险分析和对策研究

福岛核事故暴露了乏燃料水池安全研究的不足,尤其是氢气风险评价方面的不足。根据IAEA及我国相关法规要求,应对核电厂乏燃料水池发生严重事故后的氢气风险进行评估,并对氢气风险的消除进行对策研究。本文采用MELCOR程序建立分析模型,计算研究了乏燃料水池严重事故下的事故进程和氢气产生与浓度分布,评价了厂内氢气风险并定量研究了氢气风险缓解措施。分析结果表明,氢气风险是存在的。对补水、喷淋、通风和氢气复合器等缓解氢气风险措施的研究表明,注水和喷淋是可完全消除氢气风险的,但通风和氢气复合器并不能完全消除氢气风险。消除乏燃料水池严重事故下氢气风险的重点应为保证补水措施有效,对此可提高补水措施的可靠性和阻止乏燃料水池的泄漏。     

安全壳冷却对氢气风险管理的影响研究

严重事故管理(SAM)过程中,氢气控制相关的缓解措施可能与其他缓解措施相互影响,带来负面效果。本文研究了安全壳冷却应用于安全壳降压策略与氢气控制策略进行事故缓解时对氢气风险的影响。利用MATLAB开发了安全壳氢气可燃性判断辅助计算(CA)用于氢气可燃性判断。在此基础上,利用一体化分析程序建立了核电厂主系统与安全壳耦合分析模型,研究了安全壳惰化与恢复安全壳冷却对氢气风险的影响。分析表明,以50%流量开启安全壳冷却,能够维持安全壳压力且内部环境处于惰化状态,结合CA,能够通过控制安全壳压力实现缓解安全壳的氢气风险,可为技术支持中心制定相关缓解策略提供参考,提高严重事故管理导则的可执行性。     

乏燃料水池严重事故现象及管理策略研究

针对我国二代改进型三环路核电厂乏燃料水池冷却管线破口事故(LOCA)引发的严重事故,使用MECLOR1.8.6程序进行了建模计算,分析研究了严重事故进程和乏燃料组件加热、熔化以及氢气的产生等主要现象。结果表明,乏燃料水池严重事故进程相对缓慢,但乏燃料组件的熔化及产生的氢气风险还是可能最终造成放射性向环境的大量释放。此外,本文还对乏燃料水池严重事故管理导则中的应急注水策略和氢气风险管理策略的有效性进行了计算分析,得到了严重事故下执行相关策略的时间窗口,从而为同类型核电厂严重事故管理导则的开发和有效执行提供支持。     

安全壳冷却对氢气风险管理的影响研究

严重事故管理(SAM)过程中,氢气控制相关的缓解措施可能与其他缓解措施相互影响,带来负面效果。本文研究了安全壳冷却应用于安全壳降压策略与氢气控制策略进行事故缓解时对氢气风险的影响。利用MATLAB开发了安全壳氢气可燃性判断辅助计算(CA)用于氢气可燃性判断。在此基础上,利用一体化分析程序建立了核电厂主系统与安全壳耦合分析模型,研究了安全壳惰化与恢复安全壳冷却对氢气风险的影响。分析表明,以50%流量开启安全壳冷却,能够维持安全壳压力且内部环境处于惰化状态,结合CA,能够通过控制安全壳压力实现缓解安全壳的氢气风险,可为技术支持中心制定相关缓解策略提供参考,提高严重事故管理导则的可执行性。     

百万千瓦级压水堆严重事故下局部隔间氢气风险分析

核安全法规要求控制严重事故下核电厂安全壳内的氢气浓度。除安全壳整体外,局部隔间的氢气浓度同样是关注的重点。本文采用一体化严重事故分析程序对百万千瓦级压水堆核电厂安全壳局部隔间进行建模,分析了不同事故下的氢气风险。结果表明,严重事故下部分隔间短时间内可能存在燃烧风险。本文对降低燃烧风险的方法进行分析计算和筛选,得出的结论可以为安全壳隔间的设计优化提供参考依据。     

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.     

AP1000严重事故分析

AP1000设计中考虑了以下几类严重事故：堆芯和混凝土相互反应;高压熔堆;氢气燃烧和爆炸;蒸汽爆炸;安全壳超压;安全壳旁通。本工作给出了AP1000在设计时对严重事故的考虑和发生严重事故后的最终结果。     

安全壳内氢气量和超压的控制

前言　核电站严重事故中，放射性物质将向空气中散发。严重事故是造成安全壳内氢气积聚和安全壳超压的源项。安全壳内氢气的积聚和安全壳超压对安全壳的完整性和功能存在着挑战和威胁。     

严重事故后安全壳氢气风险控制论证

核安全法规要求控制严重事故下核电厂安全壳内的氢气风险.在福清核电5、6号机组的设计中,针对严重事故后氢气风险的预防和缓解采取了多项措施,包括非能动氢气复合器、预防氢气局部积聚的工程改进等.采用一体化严重事故分析程序对上述措施有效性进行了计算论证,结果表明,福清核电5、6号机组的氢气风险控制措施能够有效应对威胁安全壳完整...     

失水事故后燃料厂房空气自然循环及氢气分布实验研究

针对乏燃料水池失水、乏燃料裸露的事故情景,通过实验研究了燃料厂房内空气自然循环及氢气分布的基本规律,得到了空间气体温度场及浓度分布实验数据。结果表明,由于空气自然循环的热量输运作用,各区域升温速率随温度的升高而降低;与外界自然对流通风的简单方式可显著减缓温度上升,实验24 h后加热元件温度低于490 ℃;喷淋则可迅速降低所覆盖区域温度,空间气体平均温度在喷淋后5 min内下降了100 ℃。氢气在空气环境中将形成浓度分层,喷淋搅混可在2 min内破坏该分层;对于氢气/蒸汽混合气体,随着蒸汽凝结及外界空气涌入逐渐具有可燃性,喷淋对此无积极作用。基于上述研究结果,提出了能动和非能动的事故缓解措施建议。     

CANDU6核电厂无过滤安全壳通风模式的研究

CANDU6核电厂早期设计未考虑严重事故对策,在严重事故下,CANDU6核电厂的安全壳容易失效。为了解决这一问题,本文研究了无过滤安全壳通风模式对CANDU6核电厂安全壳的影响。本文选取典型的全厂断电严重事故,利用重水蒸气回收系统作为无过滤安全壳通风的路径,初步研究了该通风模式下对安全壳完整性的保持和对裂变产物源项的滞留能力。研究表明:该通风模式可以有效保持安全壳的完整性,同时,对裂变产物源项也有一定的滞留能力。     

A study of risk reduction for some mitigation features for PWRs

The risk reduction attainable with mitigation features in a large-dry pressurized water nuclear reactor (PWR) is evaluated. The calculations are made in a probabilistic risk analysis framework, and they are based on Zion Probabilistic Safety Study (ZPSS). Some of the modifications made to this study are also taken into account.The mitigation designs considered consist of features for simultaneously controlling late containment overpressure, containment basemat penetration, and hydrogen burning. The individual mitigation features include: a passive containment heat removal system (PCHRS), a filtered-vented containment system (FVCS), a core ladle, and controlled hydrogen burning. Emphasis is placed on comparison of PCHRS and FVCS design options. The results include calculations of the sensitivity to several failure mode probabilities and to the probability of core meltdowns with containment bypass.     

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.     

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.     

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。     

乏燃料贮存舱通风方式对气载放射性排出的数值模拟研究

In order to protect the workers in the spent fuel storage compartment from internal radiation, the airborne radioactive concentration in the spent fuel storage compartment needs to be controlled. Exhaustion of the airborne radiation is mainly realized through the ventilation system. According to the characteristics of the spent fuel storage tank, four ventilation methods are designed in this paper. Airpak software is used to simulate the four ventilation methods of the spent fuel storage tank. By comparing and analyzing the polluted steam concentration field, flow field and polluted steam trajectory, the effects of four ventilation methods on the airborne radioactive exclusion are studied. The results show that the layered air supply method Ⅱ has a better effect on the airborne radioactive emissions, and under this ventilation mode, the airborne radioactive concentration on the personnel working platform is lower than that in other three ventilation modes.     

乏燃料贮存舱通风方式对气载放射性排出的数值模拟研究

为保障乏燃料贮存舱内作业人员免受内照射伤害，需控制乏燃料贮存舱内的气载放射性浓度，气载放射性的排出主要通过通风系统实现。本文针对乏燃料贮存舱的特点设计了4种通风方式，利用Airpak软件对乏燃料贮存舱4种通风方式进行了模拟仿真，通过对比分析污染蒸汽浓度场、流场以及污染蒸汽轨迹图，研究4种通风方式对气载放射性排出的影响:结果表明，分层送风方式Ⅱ对气载放射性排出效果较好，并且在此通风方式下，人员作 业平台上的气载放射性浓度较其他3种通风方式低。      

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.