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。     

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。     

用MELCOR程序分析600 MWe核电厂乏燃料水池失去厂内外电源严重事故

利用MELCOR程序建立了600 MWe核电厂乏燃料水池计算模型,分别计算了在正常储存、正常换料和反应堆事故工况下,乏燃料水池失去厂内外电源严重事故序列。计算结果表明,燃料组件大约裸露一半后,锆水反应导致燃料熔化并产生大量氢气。分析了喷淋和注水对乏燃料水池事故的影响,分析结果表明,在燃料包壳失效前,以沸腾蒸发速率注水或喷淋能中止事故发展,并能使乏燃料水池水位缓慢回升。     

基于ASTEC程序的乏燃料水池严重事故研究

为了更深入地理解核电厂乏燃料水池严重事故的进程和后果,基于ASTEC程序建立了某型三代核电机组乏燃料水池严重事故分析模型,根据该模型,分别对正常运行,正常换料和异常换料三种不同运行状态下的长期丧失电源(SBO)和长期丧失电源叠加冷却管线破口(SBO+LOCA)所导致的严重事故进行了分析.分析结果表明,乏燃料水池事故进程...     

注水时机对压水堆严重事故进程的影响

以典型的3环路压水堆为参考对象,建立了详细的严重事故计算模型。选择一回路热段当量直径为18 cm的失水事故(LOCA)作为初始事件,采用RELAP5/SCDAP/MOD3.2为分析工具,对无注水、无缓解措施下的基准事故进程进行计算分析,研究3种不同注水时机对严重事故进程的影响。3种注水时机分别为堆芯表面峰值温度达到1100 K、1300 K、1500 K时开始注水。计算结果显示,压水堆严重事故进程对于注水的时机非常敏感。较早阶段的注水对于阻止堆芯熔化十分有效,注水较晚会恶化事故进程,加速堆芯熔化。     

某1000 MW压水堆核电厂乏燃料水池扩容热工冷却能力验证

针对国内某1000MW压水堆核电厂乏燃料水池扩容项目,使用计算流体力学(CFD)和理论分析方法,验证了扩容后的乏燃料水池热工冷却能力。在乏燃料水池至少存在一列反应堆水池和乏燃料水池冷却和处理系统(PTR)运行的冷却工况下,乏燃料水池平均水温均满足相应的验收准则,局部最高水温和燃料包壳最高温度均低于当地水的饱和温度。在2列PTR系统均失效的失去冷却工况下,计算出了乏燃料水池平均水温加热到沸腾温度的时间和燃料格架裸露的时间,为运行干预提供了指导。     

模块式小型堆乏燃料水池冷却系统设计

在福岛核电站事故后,乏燃料贮存安全的重要性得到了广泛重视,业界根据福岛核电站事故的教训,加强了相关研究。多用途模块式小型堆示范工程吸收了福岛核电站事故的经验反馈,在保证乏燃料贮存安全性的同时,兼顾提高模块式小型堆的经济性,在其乏燃料水池冷却系统设计时结合了其他堆型乏燃料水池系统的设计优点。本文从系统调研入手,通过归纳总结三代核电机组乏燃料水池冷却系统的配置特点,研究模块化小型堆的乏燃料水池冷却系统设计方案,并通过使用Flowmaster软件模拟各个工况下乏燃料水池冷却系统的流体特性,对现有的布置条件和设备选型进行校核计算,并基于计算得到的流体参数确定各工况下限流孔板的特征参数和主要工作泵的工况参数等,为设备的设计和采购提供了依据。     

用RETRAN程序进行乏燃料元件贮存水池的热工水力安全分析

开发了对核电厂乏燃料贮存水池进行热工水力分析的ＲＥＴＲＡＮ模型,按照最大热功率工况,即在乏燃料贮存水池中装满乏燃料组件（其中包括换料期间刚卸出的全堆芯燃料组件）的条件下用ＲＥＴＲＡＮ模型来评估乏燃料贮存水池冷却系统的冷却能力,并进行了几个假想方案的瞬态计算和校对计算。利用ＲＥＴＲＡＮ模型来评估乏燃料贮存水池稳态和瞬态的热工水力安全分析既方便,又精确,还可用于申请许可证的计算和估算水池的温度分布。     

乏燃料操作大厅的剂量场随屏蔽水位的变化研究

在乏燃料水池完全丧失冷却能力和补水的事故工况下,压水堆核电厂乏燃料操作大厅内的剂量率将随着乏燃料水池水位的降低逐渐升高。本文以一典型压水堆核电厂的乏燃料水池为研究对象,采用QAD-CGGP程序,计算并分析了乏燃料操作大厅内的剂量场分布及其随水位的变化规律。计算结果表明:(1)在3.786～7.736 m水层厚度范围内,操作平台处的剂量率随水层厚度的变化不明显;(2)乏燃料水池上方的剂量率峰值位于高密格架区域上方;(3)在3.436～4.736 m水层厚度范围内,乏燃料水池上方的剂量率峰值在0.914～288 μSv/h范围内变化,并随着屏蔽水层厚度的减小呈指数递增趋势,且操作平台处剂量点的剂量率均满足乏燃料操作大厅辐射分区要求;(4)满足乏燃料操作大厅辐射分区要求所需的最低水位为+15.77 m。     

乏燃料水池丧失冷却事故下安全性能评估

以RELAP/MOD3为分析工具,对典型沸水堆核电厂乏燃料水池热工水力行为进行模拟,详细分析乏燃料水池自然循环对流换热、丧失冷却性能下燃料裸露过程、应急洒水喷淋、热辐射等。验证所建立的乏燃料水池模型计算乏燃料水池冷却系统正常运行下的稳态过程可用后,对丧失冷却事故条件下的乏燃料水池丧失冷却事故下安全性能进行分析。计算结果为乏燃料水池冷却丧失性能后17.87 d乏燃料将裸露;若考虑辐射传热因素则包壳峰值温度达到1204℃的时间延后8.97 h;若按照美国核能研究所(NEI)建议的12.6kg/s喷淋洒水量,需要2.4 h可将燃料温度由726.9℃降至100℃。     

AP1000外部灾害情形下乏燃料池缓解策略研究

日本福岛核事故后,乏燃料池（SFP）在事故中的安全性得到广泛的关注。AP1000乏燃料池冷却系统（SFS）是一非安全相关的系统,不需在事故后运行以缓解设计基准事故。但乏燃料池在超设计基准事故或外部灾害事件（包括自然灾害和人为事件）下的安全性一直是核电厂设计的重点。本工作结合美国核能研究所（NEI）给出的扩大损害的缓解导则（EDMG）提出了针对AP1000外部灾害情形下的SFP缓解策略（包括内部策略和外部策略）,并对策略进行了评估。本工作结论有助于AP1000 SFP EDMG的建立,对AP1000核电厂的设计、建造、运行管理和事故管理均有很强的参考价值。     

AP1000乏燃料水池失冷瞬态特性研究

以AP1000堆型为参考，建立乏燃料水池冷却系统主要设备的热平衡耦合数学模型，并研究各类失冷事故下乏燃料水池水温的瞬态变化。模拟结果显示，在整堆芯卸料时即发生丧失所有冷却途径的事故，则燃料裸露时间约为24 h；在装料后即发生丧失所有冷却途径的事故，则燃料裸露时间约为213 h。这些工况的模拟结果为应对相应乏燃料水池失冷事故提供了参考反应时间。      

Thermal-hydraulic design and transient analysis of passive cooling system for CPR1000 spent fuel storage pool

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool(SFP). Our design can effectively manage the SFP temperature not to exceed80 C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout(SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively.The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 C after losing the heat sink.     

Beyond design basis accidents in spent-fuel pools — Generic Issue 82

Nuclear reactor plants include storage facilities for the wet storage of spent-fuel assemblies. The safety function of the spent-fuel pool (SFP) and storage racks is to cool the spent-fuel assemblies and maintain them in a subcritical array during all credible storage conditions and to provide safe means of loading the assemblies into shipping casks.Generic Issue 82 (GI-82) relates to the concern that for a postulated accident sequence that results in the loss of water from a light-water reactor (LWR) spent-fuel storage pool, a Zircaloy cladding fire could occur and propagate to older stored fuel. This issue was identified during hearings concerning SFP reracking amendments in the late 1970s when licensees were starting to use high-density storage racks. High-density racks are used to accommodate the storage of spent fuel in SFPs at reactor sites until such time as the Department of Energy (DOE) repository is available and spent fuel can be removed from the reactor sites. Maintaining a low-density storage configuration for recently discharged spent fuel would reduce the Zircaloy cladding fire probability by an order of magnitude, but at a greater cost for additional onsite storage space.The accident sequences that could result in water loss from the SFP, including beyond design basis earthquakes, various types of seal failures and dropped shipping casks, and the Zircaloy cladding fire issues have been studied by the NRC staff. The results of these studies are provided in NUREG-1353, “Regulatory Analysis for the Resolution of Generic Issue 82, Beyond Design Basis Accidents in Spent-Fuel Pools”. Although these studies conclude that most of the spent-fuel pool risk is derived from beyond design basis earthquakes, this risk is not greater than the risk from core damage accidents due to these beyond design basis earthquakes. Therefore, reducing the risk from spent-fuel pools due to events beyond the safe shutdown earthquake would still leave a comparable risk due to core damage accidents. The risk due to beyond design basis accidents in spent-fuel pools, while not negligible, is sufficiently low that the added cost involved with further risk reduction is not warranted.     

有热管冷却的乏燃料池自然对流换热特性分析

将分离式热管作为长期非能动冷却系统应用CAP1400乏燃料池,分离式热管的蒸发端布置在乏燃料池四周。本文运用数值模拟方法对具有热管冷却的乏燃料池内温度场和流场特性进行数值分析,并研究布置在池内的各排蒸发管管外对流换热强度。研究表明：当能动型冷却系统停止工作后,仅靠该非能动冷却系统可成功带走池内衰变热并保证池内不沸腾;内排蒸发管束外侧的对流换热系数高于外排蒸发管束,可达到外排管束的1.05倍,蒸发管上、下端的对流换热系数较大,中间段对流换热系数最小。研究结果对分离式热管运用于乏燃料池具有一定参考意义。     

Study on modeling of spray cooling for spent fuel pool accidents

In Japan, spray equipment is prepared in spent fuel pools (SFP) in accordance with the regulatory requirements to mitigate fuel damage in the event that the water level of SFP cannot be maintained. In order to evaluate the spray coolability of fuel assemblies in SFP accidents, the spray cooling experiments were conducted under the SFP conditions. The experimental facility contains one mock-up BWR fuel assembly with full-length 7 × 7 heater rods in a mock-up SFP rack. The measured surface temperatures indicate that the spray injection results in the top-down quench and the precursory cooling, which are consistent with the spray-cooling mechanism that has been revealed by previous studies investigating reactor core spray. Further, the numerical simulations of the experiments were conducted using the TRACE code to examine the applicability of system codes for evaluating the spray coolability of SFPs. Although the TRACE calculation with a simple analytical model reproduced the top-down quench by spray injection as observed in the experiments, some qualitative differences were found between the experiments and calculations. The causes of these differences were revealed and the applicability of system codes were discussed.     

非能动压水堆核电厂乏燃料池风险评价

以非能动压水堆核电厂为研究对象,对可能引起乏燃料损伤的内部事件进行了风险评价。采用PSA软件RiskSpectrum建立事件树和故障树模型,进行乏燃料损伤频率（FDF）定量化。结果表明：在所有工况下总的FDF为2.05×10^-9/(堆•年),远小于堆芯的损伤频率（约2.41×10^-7/(堆•年)）;即使在放射性完全释放的假设下,乏燃料损伤导致的大量放射性释放频率仍较堆芯损伤导致的大量放射性释放频率（约2.38×10^-8/(堆•年)）低1个量级;由于非能动压水堆核电厂有多重预防缓解措施以应对乏燃料池（SFP）事故,SFP风险远低于堆芯风险,可实现核安全导则的安全目标。