池式钠冷快堆放射性释放风险概率安全评价事件树分析

池式钠冷快堆的安全特性和放射性释放机制与压水堆有着显著不同,在核安全新要求下,亟待开展放射性释放风险概率安全评价(PSA)研究。本文以池式钠冷快堆为研究对象,通过分析放射性来源、包容边界及破坏包容边界完整性的严重事故现象,确定了池式钠冷快堆大量放射性释放的主要位置和释放模式,构建分析了放射性释放事件树。本文分析结果可为进一步开展池式钠冷快堆放射性释放风险PSA提供参考。     

池式钠冷快堆双层容器泄漏事故序列研究

钠冷快堆是第4代反应堆中的优选堆型,具有安全性高的特点。池式钠冷快堆的双层容器泄漏会导致一回路钠泄漏并发生严重事故。本文采用概率安全分析方法分析池式钠冷快堆双层容器泄漏事故,包括事故的确定论分析及放射性释放路径分析以及池式钠冷快堆双层容器泄漏的事故序列及定量化。结果表明,池式钠冷快堆双层容器泄漏事故后正常通风开启情况下可能发生大量放射性释放。双层容器泄漏导致的大量放射性释放频率为1.07×10^-11（堆•年）^-1,双层容器泄漏事故中大量放射性释放占比为0.1%。     

池式钠冷快堆事故源项计算方法研究及其应用

针对传统轻水堆事故源项计算方法不适用池式钠冷快堆的问题,分析可能发生的设计基准事故和超设计基准事故的释放路径,研究建立适用于池式钠冷快堆的堆芯损伤类、泄漏类和钠火类事故源项计算方法。结合示范快堆的6种典型事故：1盒燃料组件瞬时全部堵塞事故、反应堆堆本体覆盖气体边界泄漏事故、一次氩气衰变罐破损事故、主容器泄漏事故、一回路外无保护套管的钠净化管道泄漏事故和一回路无保护套管的外辅助管断裂或泄漏合并隔离阀关不住事故,开展事故源项计算及其剂量后果评价。结果表明：6种事故的放射性后果均低于GB 6249-2011的要求。该方法还可为回路式钠冷快堆、铅铋快堆以及气冷快堆事故源项计算提供参考。     

池式钠火事故下燃烧产物气溶胶行为研究

在钠冷快堆的安全评估中,分析钠泄露导致的池式钠火事故下燃烧产物的气溶胶行为尤为重要。本文采用将池式钠火燃烧模型与气溶胶动力学模型耦合的方式,开发了池式钠火事故下燃烧产物气溶胶行为分析程序REBAC-SFR,基于该程序模拟了SAPFIRE-D1和ABCOVE池式钠火实验,并与实验数据进行了对比。结果表明,本文开发的程序具有良好的可靠性和正确性,可为钠工艺间内池式钠火事故下燃烧产物气溶胶行为分析研究提供理论工具。      

CONTAIN-LMR程序中池式钠火事故分析计算模型的验证

CONTAIN-LMR是针对以液态钠为冷却剂的反应堆而开发的安全壳事故一体化分析程序。我国目前的CONTAIN-LMR程序版本为2000年左右从法国引进,还未进行过面向工程设计的系统性地程序开发和验证。本文主要针对CONTAIN-LMR程序中模拟池式钠火事故的分析模型进行详细分析,并采用国际上的池式钠火实验进行验证,实验验证结果表明CONTAIN-LMR程序可以较准确地模拟池式钠火事故造成的钠工艺间内的温度、压力升高及放射性钠气溶胶行为。本文的研究结果初步表明CONTAIN-LMR程序可用于钠冷快堆的钠火事故分析。     

池式钠冷快堆熔融物堆内滞留初步分析研究

为防止堆芯熔毁后熔融物熔穿反应堆容器,造成大量放射性释放,三、四代反应堆设计中普遍考虑了熔融物滞留方案。池式钠冷快堆在主容器底部安装堆芯熔化收集器,对熔融物进行有效收集和长时冷却。利用中国原子能科学研究院自主开发的FRTAC程序,计算堆芯熔毁后主容器内的自然循环,分析熔融物长时冷却过程,研究钠冷快堆的熔融物堆内滞留方案。结果表明：熔融物掉落至堆芯熔化收集器上后,主容器内的自然循环可以有效冷却熔融物,并由事故余热排出系统将余热导出至大气环境中。     

钠冷快堆混合钠火程序开发

钠管道泄漏继而发生钠的燃烧为钠冷快堆特有的事故。在喷雾钠火模型和池式钠火模型基础上,将钠喷雾燃烧和池式燃烧进行了耦合,并针对钠冷快堆钠工艺间的结构特点,最终开发了混合钠火计算程序COMSFIRE。使用该程序计算了FAUNA喷雾钠火试验和CADARACHE池式钠火试验,并与试验结果和部分程序计算结果进行了对比。同时设计了混合燃烧算例,并使用该程序与CONTAIN-LMR程序进行了对比。通过计算结果的对比和分析,初步验证了程序的正确性。     

钠冷快堆载热系统设计

本文概述钠冷快堆的载热系统,综合快堆冷却系统的设计原则和方法,提供某些重要数据。对钠冷快堆一次钠回路系统的“池式”和“管式”布置方案作了分析比较,认为实验快堆采用“管式”双回路载热系统比较合适。     

池式钠冷快堆非能动停堆技术方案研究

非能动停堆技术一直是快堆安全技术研究的热点,受到了国际上研究快堆技术国家的重视。目前,国内也开展了相关技术研究。本文在调研和分析的基础上,从非能动停堆技术的安全特性、技术成熟度和中国实验快堆(CEFR)典型事故的分析等几个方面进行了两种有一定试验经验的非能动停堆装置的对比,并给出了中国池式钠冷快堆非能动停堆技术发展的建议。     

应用于钠冷快堆的超临界二氧化碳动力转换系统研究

超临界二氧化碳（SCO₂）布雷顿循环由于高效、紧凑和可避免钠水反应等特性而成为钠冷快堆的理想动力转换系统。本文以1 200 MWe大型池式钠冷快堆为系统热源,钠回路温度及热负荷为循环系统运行边界,对比研究了不同SCO₂布雷顿循环系统性能和关键设备性能的变化规律。研究发现,级间冷却再压缩循环与钠冷快堆热源特性匹配性最佳,且循环效率最高（40.7%）。进而研究了不同运行参数对级间冷却再压缩循环效率的影响规律,给出了循环系统效率对各关键影响因素的敏感度,发现循环系统效率对冷端参数的敏感度最强,其次为分流比和透平入口参数,对主压缩机级间压比的敏感度最弱。     

Level-1, -2 and -3 PSA for AHWR

In nuclear reactor probabilistic safety analyses (PSAs), risk is usually defined by the frequency and magnitude of radioactive releases to the environment (Generic CANDU, 2002). An integrated Level-1, -2 and -3 PSA have been carried out for thorium based natural circulation driven advanced heavy water reactor (AHWR). A Level-1 PSA models accident sequences up to the point at which the reactor core either reaches a stable condition or becomes severely damaged, releasing large amounts of radionuclides into the containment. The probabilistic aspects of the analysis focus on the performance and reliability of nuclear plant systems and station staff in response to plant upsets. A Level-2 PSA examines severe reactor accidents through a combination of probabilistic and deterministic approaches, in order to determine the release of radionuclides from containment, including the physical processes that are involved in the loss of structural integrity of the reactor core (Generic CANDU, 2002). A Level-3 PSA goes through the short and long term (radiological) effects on the public (Fullwood, 2000). In this study the risk associated with internal events is only addressed. In the first phase, Level-1 PSA has been carried out to identify postulated initiating events (PIEs) which may lead to severe core damage (SCD) for the reactor. In the second phase, a Level-2 PSA examines two enveloping severe accidents through a combination of probabilistic and deterministic approaches and determines the release of radionuclides from containment. In the third phase, a Level-3 PSA is carried out for the transport of radionuclides through the environment and for the evaluation of public health risk for the two scenarios considered. The salient findings are presented in the paper.     

西安脉冲堆概率安全分析技术要点及分析框架研究

针对西安脉冲堆（XAPR）自身设计特点及安全特性,研究了XAPR概率安全分析（PSA）的技术特殊要点,提出了XAPR PSA分析框架及技术要素具体实施方法。最后以XAPR堆水池中破口失水事故为始发事件,验证了XAPR PSA研究思路。分析表明:以始发事件为起点、事件序列为主干、放射性释放类为终点的一体化事件树结构分析框架适合于XAPR PSA。      

Prevention and mitigation measures to ensure containment integrity

The consequences of severe reactor accidents depend greatly on containment safety features and containment performance in retaining radioactive material. In most severe accident sequences, the ability of the containment boundary to maintain its integrity is determined by two factors: (1) the magnitude of the loads; and (2) the response to those loads of the containment structure and the penetrations through the containment boundary. Severe accident phenomenology and consequences has been the subject of worldwide intense research for more than 20 years. In this paper, the main threats to the containment integrity are reviewed, the state of knowledge and remaining uncertainties are briefly described, and the links with the relevant Community research activities are made. Finally, areas in which further research may be needed are listed.     

大型先进压水堆IVR策略评估和裂变产物分布计算

大型先进压水堆通过堆内熔融物滞留（IVR）策略来缓解严重事故后果以降低安全壳失效风险。其中堆腔注水系统（CIS）被引入来实现IVR。本文使用严重事故分析软件计算大型先进压水堆在冷管段双端断裂事故下的事故进程、热工水力行为、堆芯退化过程和下封头熔融池传热行为,评估能动CIS的事故缓解能力。计算结果表明,事故后72 h,下封头外表面热流密度始终低于临界热流密度（CHF）,表明IVR策略有效。此外,计算分析了惰性气体、非挥发性和挥发性裂变产物的释放和迁移行为。计算发现,IVR下更多的放射性裂变产物分布在主系统内,壁面核素再悬浮形成气溶胶的行为被消除,安全壳壁面上沉积的核素被大量冷凝水冲刷进入底部水池。总体来说,IVR策略能更好地管理放射性核素分布,减小放射性泄漏威胁。     

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

The accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10^?5 and 1×10^?6 per plant operating year, respectively, which were selected based on the IAEA's safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.     

应急设施可居留性分析的严重事故源项初步探讨

以先进压水堆核电厂为对象,开展了适用于应急设施可居留性评价的严重事故源项分析方案研究,覆盖了堆芯释放、安全壳内自然去除、放射性物质向环境释放途径等。结合非能动安全壳冷却系统的特征,重点研究了安全壳可能的失效行为,论证了安全壳在事故后24h和72h失效工况的辐射影响。结果表明:两种工况放射性释放水平均达到了INES(国际核事件分级)第6级的水平,属于比较严重的核事故;133 Xe、131I为主导核素组的主导核素,所释放的133 Xe介于WASH-1400中PWR2~PWR4之间的水平,131I介于PWR5~PWR6之间水平。同时,以国内某沿海厂址为例,评价了两种工况下应急指挥中心(EOF)工作人员的有效剂量,均可满足100mSv的剂量限值要求。     

Severe accident mitigation strategy for the generation II PWRs in France – Some outcomes of the on-going periodic safety review of the French 1300 MWe PWR series

The 3rd Periodic Safety Review of the French 1300 MWe PWRs series includes some modifications to increase their robustness in case of a severe accident. Their review is based on both deterministic and probabilistic approaches, keeping in mind that severe accidents frequencies and radiological consequences should be as low as reasonably practicable, severe accidents management strategies should be as safe as possible and the robustness of equipment used for severe accident management should be ensured.Consequently, the IRSN level 2 probabilistic safety assessment (L2 PSA) studies for the 1300 MWe reactors have been used to re-assess the results of the utility's L2 PSA and rank them to identify the containment failure modes contributing the most to the global risk. This ranking helped the review of plant modifications.Regarding strategies for accident management, the EDF management of water in the reactor cavity during a severe accident for the 1300 MWe PWRs is presented as well as the IRSN position on this strategy: this is an example where the optimal severe accident management strategy choice is not so easy to define.Regarding the robustness of equipment used for severe accident management, the interest of a diversification or redundancy of the French emergency filtered containment venting opening is one example among many others.This paper presents the analysis conducted by IRSN during the 3rd periodic safety review of the French 1300 MWe PWRs. Future NPP upgrades to limit radioactive releases in case of containment filtered venting, to prevent containment venting and basemat melt-through are analysed in another framework (post-Fukushima and long-term operation projects).     

Exact Solutions of Two Point Time Correlations on Point Reactor Kinetics with Gaussian Reactivity Fluctuations

Station blackout is reported to be a sequence that would likely be a significant contributor to the accident risk at a boiling water reactor (BWR). The occurrence frequency of station blackout is evaluated in probabilistic safety assessment (PSA) to be 6×10^?6 per reactor year at Limerick and less than 10^?7 per reactor year at BWR in Japan.

This report describes an analytical study of thermal-hydraulic and radionuclide behavior during a postulated severe accident of station blackout at a reference BWR plant. The analytical approach was shown in both of hand calculation and the THALES/ART code calculation to better understand wide physical and chemical phenomena in the processes of severe accidents.

We evaluated timing of key events, core cooling and core temperature, reactor vessel failure, debris temperature, containment pressure, and release and deposition of radionuclide in the containment. The THALES and CORCON models on the chemical reactions in the core-concrete interaction lead to great differences in the increasing rate of containment pressure and the release rate of fission products from the core debris.     

三代压水堆核电厂DCH参数敏感性研究

安全壳直接加热（DCH）是导致安全壳早期超压的主要贡献之一,严重威胁安全壳完整性,并可能造成放射性物质早期大量不可控释放。本文以我国某三代压水堆为研究对象,首先基于风险导向的事故分析方法（ROAAM）,利用双隔间平衡（TCE）模型编写程序计算典型事故工况下的DCH载荷;其次结合安全壳失效概率曲线得出DCH现象造成的安全壳失效概率;最后对计算程序中不易得到的参数或经验值等不确定性较大的参数进行敏感性分析,归纳敏感性分析结果,找出敏感参数的不确定因素。结果表明：熔融物质量、堆腔几何设计、安全壳布置设计会直接影响DCH后果。