MELCOR/MACCS simulation of the TMI-2 severe accident and initial recovery phases, off-site fission product release and consequences

MELCOR has become the preferred code of the Swiss nuclear industry and of PSI for severe accident analysis, on account of its integrated systems-level approach and validation against experiments and more detailed codes, while MACCS is commonly used by safety authorities for independent assessment of off-site consequences, in particular health effects. The present work arises out of a programme to assess MELCOR independently using empirical data consistent with the recommendations of the OECD/CSNI validation matrix for core degradation codes. The MELCOR 1.8.5RD calculations are based on a model for phases 1 and 2 provided by the code developers but with a simplified thermal hydraulic noding in certain regions and the inclusion of a simple representation of the fission product release and transport pathways. The model has also been extended to simulate phases 3, 4, and the continuing initial period of core recovery and stabilisation. These calculations are a first attempt to demonstrate a MELCOR–MACCS capability to simulate the whole plant accident sequence beyond phase 4, including the containment response and off-site consequences arising from fission product release from the containment. Emphasis is placed on the overall accident evolution and whole plant response, rather than the detailed behaviour. Results are compared with observed and deduced data for the major accident signatures and rough estimates for exposure based on off-site monitoring. The results provide a good basis for the NPP analysis foreseen.     

核电厂严重事故下安全壳通风导致放射性后果的快速评价

提出一种严重事故下安全壳通风导致放射性后果的快速评价方法。通过预先计算通风后安全壳的释放份额和1%初始堆芯总量释入安全壳时的公众个人终身剂量,以及通过事故下安全壳的辐射监测仪表间接得到堆芯向安全壳的释放份额,能够快速评价厂外不同距离处公众的个人终身剂量,它可为严重事故的管理和厂外应急策略的实施提供强有力的支持。     

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).     

福岛核电厂3号机组严重事故模拟分析

本文应用MELCOR程序,通过建立全厂详细的模型,对福岛第一核电厂3号机组在地震发生后3 d内的严重事故进程进行了模拟分析并与电厂实测数据进行了比较,再现了从事故开始到堆芯失效坍塌直至氢气爆炸在内的主要严重事故现象。基于文中假设的模拟计算得到的趋势与电厂现有实测数据较为一致,结果表明：地震发生后约36 h反应堆水位降至堆芯活性区顶部。操纵员未能及时成功对安全壳和反应堆进行快速卸压,以在堆芯底部出现裸露前向反应堆补充冷却水,使得堆芯出现严重的锆水反应,大部分燃料包壳已破损而导致易挥发的放射性裂变产物的释放;但此时堆芯整体依然维持可冷却几何形状;在消防水泵向反应堆注入冷却水期间,由于冷却注入流量出现中断,导致堆芯进一步熔毁坍塌;碎片迁移至下腔室后,进一步的锆水反应(金属 水反应)新增的氢气泄漏并积聚在反应堆厂房上部,引发了安全壳厂房的爆炸;72 h内,堆芯内约50%的锆合金发生了氧化,压力容器下封头未发生失效。     

Development of accident management measures for RBMK-1500 in the case of loss of long-term core cooling

Results of the Level 1 Probabilistic Safety Assessment of the Ignalina Nuclear Power Plant have shown that in the risk topography transients are dominating. Analysis has shown that failure of the long-term core cooling is the main contributor to the core damage frequency. However, the reactor core damage in the long-term indicates the potential opportunities for the accident management. The main goal of accident management is to avoid a multiple fuel channel rupture because considering the design of RBMK reactors the consequences of rupture of more than 11–16 FC at full pressure would be close to the consequences of Chernobyl accident. This paper presents a detailed thermal–hydraulic analysis of the accidents with the loss of long-term core cooling, performed using the RELAP5 model of Ignalina NPP reactor cooling circuit and safety systems. Different ways of potential accident management are discussed. On the basis of this analysis the accident management strategy was developed.     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.     

严重事故下开启双层安全壳环形空间通风过滤系统对缓解放射性向环境释放影响研究

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。     

秦山一期核电站SGTR导致堆芯熔化进程及事故缓解措施的研究

采用自行研制的核反应堆严重事故分析平台,对秦山一期核电站蒸汽发生器传热管破裂(SGTR)初因导致堆芯熔化严重事故进程进行了分析研究,并根据美国SAN ONOFRE核电站的1PE结果以及SURRY的PSA评估结果,选择适当的缓解措施,如一回路补给水、二回路补给水、一回路卸压等,对该事故做了相应的严重事故管理。通过计算分析,对阻止SGTR导致堆芯熔化进程的缓解措施的有效性进行了验证：     

Core temperature transient in the early phase of core uncovering at TMI-2

This paper describes a best-estimate analysis of the initial core boil-down and heat-up transient at Three Mile Island Unit (2) on 28 March 1979. This transient began shortly after all reactor coolant pumps were secured (100 min after reactor trip) and was terminated by a period of sustained high pressure injection of emergency cooling water, starting at 202 min.

The analysis is primarily directed to understanding the progression of core damage, rather than providing a detailed characterization of the core end-state condition. The latter objective can be achieved only after vessel head removal and visual examination.

The thrust of the present effort has been to: (1) develop a core coolant mixture level (dry-out level) calculation which satisfies the boundary conditions implied by various instrument responses and system operational characteristics; (2) couple the level calculation with a core heat-up modelto simulate the accumulation of thermal damage in the exposed, upper regions of the core; (3) compare calculated gross damage to the core with measurements of hydrogen and fission product releases subsequent to the accident.

Results indicate that:

1. (i) Observed containment hydrogen levels were due to Zircaloy/stainless steel corrosion that occurred during the period of core uncovering between the de-activation of the loop A reactor coolant pump (100 min after trip) and sustained operation of the high pressure injection system 100 min later. Appreciable zircaloy oxidation probably commenced at 150 min after trip, and continued at a high rate until the sustained high pressure injection at 202 min caused a major core quench.

2. (ii) There was some potential for fuel liquefaction. Calculations imply that peak fuel temperatures did not exceed the UO₂ pellet melting temperature, but 30% of the fuel was exposed to temperatures where liquid U---Zr---O alloys could have formed.

3. (iii) A substantial fission product release was obtained from fuel over-heating; however, an apparent disparity between the expected fission product release by calculation and the high range of fission product estimates obtained from plant measurements suggests that a significant release fraction may have originated from powdered or rubbilized fuel during cooldown. Additional gas releases may have developed from hot spots which persisted after core quench.

4. (iv) Steam temperatures in the upper plenum, at the outlet nozzle elevation, were generally below 900°C (1650°F) although this value was probably exceeded for a few min during the partial fuel quench caused by activation of the loop 2B reactor coolant pump, at 174 min after trip. The metal-work in the upper plenum, above the upper tieplate did not experience appreciable heating.

Thermal damage to the fuel and consequential weakening and mechanical disruption of the core was essentially complete 230 min after turbine trip.     

秦山一期核电站小破口冷却剂丧失初因严重事故以及缓解措施的研究

采用基于SCDAP／RELAP5的核反应堆严重事故分析平台．分析研究了秦山一期核电站一回路冷段小破口冷却剂流失(SBLOCA)初因导致严重事故进程,并根据美国SANONOFRE核电站的IPE结果以及SURRY的PSA评估结果,选择适当的缓解措施,即进行一回路补给水,对该事故做了相应的干预。通过计算分析,对阻止SBLOCA引发的严重事故进程的缓解措施的有效性进行了验证。     

Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TM1–2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating.     

蒸汽发生器完全丧失给水引发的压水堆严重事故研究

采用严重事故最佳估算程序SCDAP/RELAP5/MOD3.4,建立了美国Surry核电站的详细计算模型,对完全丧失给水(TLFW)引发的堆芯熔化事故进行了研究分析.为准确预测压力容器内堆芯熔化的进程,给二级PSA提供可信的初始条件,计算中考虑了一回路压力边界的蠕变破裂失效,并评价了人为干预对堆芯熔化进程及事故后果的影响.     

The passive ALWR approach to assuring containment integrity

Design requirements for the advanced light water reactor (ALWR) have been developed so as to provide high assurance of containment integrity even in the event of a severe accident. The containment integrity requirements are in the form of two design criteria, and associated methodology, which address containment severe accident performance and offsite dose and are specified in the ALWR utility requirements document (URD), a set of detailed design requirements for next generation plants in the US.The containment performance criterion, which is the main focus of this paper, specifies that plant design characteristics and features shall be provided to preclude core damage sequences which could bypass containment and to withstand core damage sequence loads. This containment performance capability, along with the associated dose mitigation capability, provides a technical basis for emergency planning change since there would not be the same need for rapid offsite emergency response that is called for under the existing US emergency planning basis.     

压水堆核电厂安全壳内碎片源的踏勘方法介绍

安全壳地坑是许多压水堆核电厂设计为在失水事故后为堆芯冷却和安全壳排热提供再循环水的专设安全设施。安全壳内的潜在碎片源在事故中可能堵塞安全壳内的地坑滤网,从而造成安全壳地坑性能下降。为了评价安全壳地坑在破口事故后能否满足设计要求,首先应确定潜在碎片源的类型以及它们在安全壳内的位置。安全壳内现场踏勘就是寻找与定位碎片源的有效方法,并能够提供一些进行安全壳地坑性能分析的必要信息。介绍了压水堆核电厂安全壳内碎片源的一些踏勘方法。     

压水堆核电站完全丧失给水引发的严重事故研究

采用严重事故最佳估算程序RELAP5/SCDAPSIM/MOD3.2，建立美国Surry-2核电站的详细计算模型，对完全丧失给水（TLFW）引发的堆芯熔化事故进行研究分析。为准确预测压力容器内堆芯熔化的进程，为二级概率安全评价提供可信的初始条件，计算中考虑了一回路压力边界的蠕变破裂失效，并评价了人为干预对堆芯熔化进程及事故后果的影响。计算结果表明，由完全丧失给水引发的压水堆核电站严重事故不会出现人们担心的高压熔堆；反应堆压力容器下封头的失效位置不是在其底部，而是在其侧面；通过打开稳压器释放阀对一回路实施主动卸压能够大大推迟事故的进程。     

Methodology and results of the seismic probabilistic safety assessment of Krško Nuclear Power Plant

A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Kr

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ko plant. The methodology adopted is the seismic PSA (probabilistic safety assessment). The Kr

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ko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of site hazard, calculation of plant structures response including soil–structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Relay chatter analysis and soil stability studies were also performed. The seismic PSA described here is limited to the analysis of CDF (level 1 PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Kr

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ko NPP but are not further described in this paper. The results of the seismic PSA study indicates that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable with most US and Western Europe NPPs located in high seismic areas.     

核电厂核事故应急情况下某些操作干预水平默认值

本文根据IAEA-TECDOC-955给出的核电厂核事故应急情况下操作干预水平（OIL）的计算公式和InterRAS1.3计算程序,分别计算了压水堆核电厂两种假想严重事故（堆芯熔化/安全壳完整性丧失或泄漏事故、蒸汽发生器完整性严重丧失事故）烟羽照射期间撤离和服用碘片的操作干预水平OIL1和OIL2;讨论了相关时间（例如预期烟羽照射时间、放射性开始向环境释放时间）、气象条件（风速、混合层高度、稳定度和降雨）、距事故源距离和释放方式（低架和高架释放）等对OIL1和OIL2计算值的影响.在此计算和讨论的基础上,对所假想的严重事故推荐了相应的OIL1和OIL2默认值;强调指出了OIL1和OIL2依赖于事故类型及事故放射性在释放到环境之前是否被去除减少等因素,其默认值须按照不同事故类型及事故放射性被去除减少的特征分别给出.     

SCDAP/RELAP5与MELCOR程序对堆芯损伤过程预测的比较

SCDAP／RELAP5与MELCOR程序是目前得到广泛使用的两个严重事故分析程序．它们在模拟堆芯溶化及压力容器下封头失效过程中采用了基于不同理论的计算模型。本文利用两个程序分别对秦山二期核电厂发生假想的全厂断电事故下的堆芯损伤过程进行预测．并对比分析了这2个严重事故分析程序的优点及相应的计算结果．     

Containment integrity is a major part of KKL-safety systems PRA

Assuming a hypothetical accident with core meltdown, hardware changes are described with which the containment integrity of the Leibstadt NPP can be secured. Venting of the containment through the standby gas treatment system is initiated manually. This system is called COSA (Containment Safeguard).The COSA system has been compared with other proposed European Systems and it is at least equivalent or better.The pressure margins for the containment failure are determined and compared with the design pressure.The expected population doses without and with COSA are presented.On the basis of a fault tree PRA the gain in overall safety is derived. A cost-benefit analysis on the three safety levels 1, 2 and 3 leads to a general formula as a proposal to minimize the ratio of investment cost to safety gain.     

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.