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为评价"华龙一号"核电厂严重事故下气载放射性排放控制措施的有效性和先进性,开展了"华龙一号"严重事故下气载放射性排放控制研究。首先,介绍了核电厂中放射性物质的产生及放射性物质向环境释放的4个途径。其次,阐述了放射性物质的主要去除机制,包括自然沉积、池式洗涤、过滤和喷淋等,以及各去除机制所涉及的气溶胶行为如气溶胶凝聚、气溶胶沉积和吸湿效应、碘化学反应等,和各去除机制所应用的设备或系统。然后,梳理了"华龙一号"在严重事故工况下所采用的几种放射性释放控制和管理措施,包括双层安全壳与环形空间通风系统、安全壳喷淋系统、安全壳过滤排放系统和严重事故管理导则中针对安全壳旁通释放的管理策略,并对不同措施控制放射性释放的效果进行计算分析。计算结果显示采用相关放射性释放控制措施比未采用时向环境的放射性物质释放能够降低1~3个数量级,说明"华龙一号"的设计及严重事故管理措施,能够有效减少事故下的放射性后果,从而减少气载放射性排放对公众和环境的影响。 相似文献
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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。 相似文献
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对百万千瓦级压水堆核电厂的安全壳内进行隔间,应用IRSN和GRS等联合开发的ASTEC程序计算该类型核电厂在发生蒸汽发生器完全失去给水严重事故工况下放射性裂变产物在安全壳内释放迁移的情况,给出了主要隔间内的放射性活度.根据安全壳内喷淋系统能否正常启用对各个隔间内的放射性活度进行了比较.结算结果表明,喷淋能否启用,对Xe、Kr等惰性气体在各隔间内分布几乎无影响;但可以大大降低I、Br等易生成气溶胶、水溶性较好的裂变产物的浓度.对其他主要以气溶胶形态存在于安全壳气空间中的裂变产物也有很强的去除作用.喷淋的成功启用,将大部分放射性裂变产物冲刷入下部的地坑区,使得安全壳内上部空间的放射性活度有了明显的降低,但裂变产物聚积在地坑,使地坑的活度大大提高. 相似文献
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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。 相似文献
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核电站发生严重事故后,安全壳能包容从堆芯释放出的裂变产物,防止向环境的大量释放,但即使在安全壳完好的情况下,仍然会存在一定量泄漏。目前国际上的三代核电机型,大多采用双层安全壳的设计,对裂变产物具有一定的包容、滞留和过滤作用。本文基于我国自主设计的第三代核电机组,结合双层安全壳的设计特点和特定源项分析,对严重事故下双层安全壳之间的环形空间及其通风过滤系统对缓解裂变产物向环境释放的作用进行了定量分析,结果显示双层安全壳及环形空间通风过滤系统能够显著降低放射性气溶胶对环境的释放,对惰性气体也有一定的延缓排放作用。 相似文献
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CPR1000核电站严重事故重要缓解措施与严重事故序列 总被引:2,自引:0,他引:2
CPR1000核电站采用非能动氢气复合器、稳压器卸压功能延伸以及安全壳卸压过滤排放系统作为严重事故的预防和缓解措施,保证在严重事故条件下核电站安全壳的完整性不受损坏,保护环境周围的居民不受核辐射的危害。通过相关严重事故谱分析,选取冷却剂管道热段双段断裂+失去应急堆芯冷却系统、全厂断电、主蒸汽管道断裂+失去喷淋、失水未能紧急停堆的预计瞬态(ATWS)这4种严重事故作为CPR1000核电站的重要严重事故序列,包络了所有安全壳内氢气产生速度快浓度高、安全壳超压、冷却剂系统发生高压熔堆、反应堆不能停堆等最严重的事故。 相似文献
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采用严重事故一体化分析程序MELCOR,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进行校核计算研究,获得了严重事故工况下核电厂关键参数的瞬态特性和非能动系统响应特性,并与安全分析报告中MAAP的计算结果进行了对比分析。结果表明:虽然校核计算结果与安全分析报告中的结果存在一定差异,但总体上事故序列和主要参数的变化趋势吻合良好,并且都能够在严重事故情况下保持压力容器和安全壳的完整性,放射性裂变产物释放量极低,缓解措施的设计能够有效缓解事故进程,满足核电厂的安全要求。 相似文献
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为了确保有效的缓解严重事故,需要对用于缓解和监测严重事故进程的重要设备、仪表在严重事故环境下的可用性进行评估.而温度、压力、湿度、辐射等参数是可用性评估的重要输入条件.本文针对百万千瓦级压水堆核电机组,参考美国核管会发布的《轻水堆核电厂事故源项》(NUREG-1465)关于严重事故后放射性物质的释放阶段和释放份额的假设,计算出事故后由堆芯释放到安全壳内的放射性源项.对于放射性物质在安全壳内的分布,不考虑喷淋和泄漏的影响,计算并分析了严重事故后安全壳内的γ和β辐射环境条件,并与APl000的设备鉴定源项进行了对比分析.本文的计算对于设备和仪表在严重事故后的可用性分析以及其所需耐受的辐射条件具有重要的参考意义. 相似文献
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The fire spray system (FSS) of the Advanced Passive PWR, as a part of the fire protection system, can provide a non-safety related containment spraying function for severe accident mitigation which is included in the Severe Accident Management Guidelines (SAMG) of the Advanced Passive PWR when dealing with severe accidents. The effectiveness of the FSS is investigated on three effects for severe accident mitigation which are controlling the containment condition, washing out fission product and injecting into the containment through three representative severe accident scenarios analysis with integral accident analysis code since there is no sufficient data support, besides the negative impact is also discussed. Results show that the FSS can be effective for controlling the containment condition, washing out fission product and injecting into the containment, however the effect is limited due to system limitation: the FSS can only cool the containment atmosphere for a short term; the flow rate of FSS cannot fulfill the success criteria given in the PRA report of the Advanced Passive PWR. Meanwhile, the hydrogen concentration and the containment water level should be the long-term monitored because actuating the FSS may cause hydrogen risk in the containment and containment flooding. Despite its limitation and negative impact, the FSS can be effective as an alternative severe accident mitigation measurement for postponing the process of accidents for safety system recovery. 相似文献
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大型干式安全壳消氢系统的初步设计 总被引:1,自引:0,他引:1
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. 相似文献
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All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by secondary containments. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident.The fission product retention capability of an intact secondary containment will depend on several factors. Recent analyses indicate that the major factors influencing secondary containment effectiveness include: the mode and location of the primary containment failure, the internal architectural design of the secondary containment, the design of the standby gas treatment system, and the ability of fire protection system sprays to remove suspended aerosols from the the secondary containment atmosphere. Each of these factors interact in a very complex manner to determine secondary containment severe accident mitigation performance.This paper presents a brief overview of US BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented. 相似文献
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Wonjun Choi Hwan-Yeol Kim Rae-Joon Park 《Journal of Nuclear Science and Technology》2017,54(3):337-347
During a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects. 相似文献
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M. Tiltmann 《Nuclear Engineering and Design》1985,90(2):161-165
For a large nuclear power plant under normal operating conditions a leakage rate for the containment of 0.25 vol.%/day is admissible. During a successfully controlled LOCA leakages of the containment will be released through filters by the annulus* air exhausting system into the environment. During a core melt accident a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. When openings in the containment steel shell occur before a catastrophic failure, a depressurization into the annulus takes place. The area of the openings determines the depressurization rate and the thermodynamic conditions in the annulus. Furthermore the behaviour of the components being necessary for accident mitigation is influenced too. This paper discusses the thermodynamic consequences of leaks in the containment shell of a German PWR during a core melt accident. The results of those calculations are the necessary boundary condition for the estimation of fission product retention in the annulus. 相似文献
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It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS. 相似文献
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