Severe accident analysis for a typical PWR using the MELCOR code

Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident.     

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。     

百万千瓦级压水堆严重事故后再注水的有效性评价

根据现有的设计资料,使用一体化严重事故分析程序MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10kg/s)、中流量(50kg/s)和大流量(200kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。     

Effect of water injection on hydrogen generation during severe accident in PWR

Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied.The analyses were carried out with different water injection rates at different core damage stages.The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region.Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K,because the core is quenched and reflooded quickly.The water injection at the peak core temperature of 1900 K,the hydrogen generation rate increases at low injection rates of the water,as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate.At peak core temperature of 2100-2300 K,the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core.Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture.Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region.However,hydrogen is generated if water is injected into the molten pool,because steam serves to the crust supporting the molten pool.Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation.Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.     

Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

This study focuses on the in-vessel phase of severe accident management (SAM) strategy for a hypothetical 1000 MWe pressurized water reactor (PWR). To examine the effectiveness of SAM strategy, it is necessary to identify and assess epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is reactor coolant system (RCS) depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, sensitivity analyses are carried out to assess the impact of the cutoff porosity related to the flow area of core node (EPSCUT), the critical temperature (TCLMAX) and the minimum fraction of oxidized Zr (FZORUP) for cladding rupture, and the flag to divert gas flows in the core to the bypass channel (FGBYPA) on the core melting and relocation process. In this study, the effect of injection time on the integrity of RPV has been examined based on the quantification of the scenario and phenomenological uncertainties.     

Intercomparison of results for a PWR rod ejection accident

This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the United States, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general, the agreement between methods was very good, providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.     

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.     

百万千瓦级压水堆核电厂低温水密实超压保护改进

针对百万千瓦级压水堆核电厂低温水密实超压保护提出改进方案,即在低温水密实状态下调低稳压器安全阀的开启/关闭压力整定值,由稳压器安全阀和余热排出系统(RRA)安全阀一起对反应堆冷却剂系统(RCP)提供双重的低温超压保护。RRA正常运行时由RRA安全阀提供超压保护,如果RRA安全阀因隔离而不可用,则由稳压器安全阀提供后备的超压保护。分析结果表明,稳压器安全阀可以在低温水密实状态下对RCP提供有效的超压保护,从而确保RCP压力边界的完整性。     

Thermal hydraulic performance of a moisture separator reheater at the 900 MWe PWR Saint Laurent Unit B1

A FORTRAN computer code is written to model the thermal hydraulic performance and behavior of a moisture separator reheater (MSR) at the 900 MWe PWR Saint Laurent Unit B1. MSR unit heat transfer coefficients, axial variations of steam quality, void fraction and local heat transfer coefficients in the MSR superheater sub-bundles are computed. Agreement between the global numerical results and measurements collected by Electricité de France (EDF) varied between 7.3% and 12%. No axial data were available for comparison. Correlations for MSR unit and sub-bundle overall heat transfer coefficients as functions of steady-state electrical power are given. The temperature response time of the MSR tube walls during transition from wavy/stratified to plug/slug flow is determined to be on the order of half a second.     

压水堆核电厂严重事故对策

描述了严重事故的过程和现象，分析了严重事故管理。系统地介绍了西屋用户集团（WOG）严重事故管理技术基础和构成：严重事故管理导则（SAMG）的主控室导则、技术支持中心（TSC）使用导则、计算辅助导则和退出导则。归纳了西屋事故对策的整体逻辑，并对我国开展严重事故对策研究提出建议。     

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。     

小型压水堆严重事故下一回路承压管道蠕变预测分析模型开发

一回路承压管道蠕变是压水堆核电厂严重事故重要现象之一。针对小型压水堆,本文基于SCDAP/RELAP5程序开发了严重事故分析模型,利用实验拟合方法得到了一回路主管道(SA321)、自然循环式蒸汽发生器传热管(00Cr25Ni35Al Ti)两种材料蠕变预测分析模型,改进了SCDAP/RELAP5程序蠕变预测分析功能模块,并通过假想事故序列验证了SA321、00Cr25Ni35Al Ti蠕变预测分析模型的合理性。为后续开展小型压水堆严重事故下一回路承压管道蠕变规律研究提供基础参考。     

Thermal-hydraulic analysis on the effect of operator action in the loss of shutdown cooling system accident in a PWR

It is necessary to develop PSA methodology and integrated accident management technology during low power/shutdown operations. To develop this technology, thermal-hydraulic analysis is necessarily required to access the trend of plant process parameters and operator's grace time after initiation of the accident. In this study, the thermal-hydraulic behavior in the loss of shutdown cooling system accident during low power/shutdown operations at the Korean standard nuclear power plant was analyzed using the best-estimate thermal-hydraulic analysis code, MARS2.1. The effects of operator's action and initiation of accident mitigation system, such as safety injection and gravity feed on mitigation of the accident during shutdown operations are also analyzed.When steam generators are unavailable or vent paths with large cross-sectional area are open in the accident, the core damage occurs earlier than the cases of steam generators available or vent paths with small cross-sectional area. If an operator takes an action to mitigate the accident, the accident can be mitigated considerably. To mitigate the accident, high-pressure safety injection is more effective in POS4B and gravity feed is more effective in POS5. The results of this study can contribute to the plant safety improvement because those can provide the time for an operator to take an action to mitigate the accident by providing quantitative time of core damage. The results of this study can also provide information in developing operating procedure and accident management technology.     

Investigation of PWR accident situations at BETHSY facility

For several years an extensive programme of separate effect tests related to PWR safety has been conducted in France and in particular at the Nuclear Center of Grenoble. Recently the BETHSY integral test facility - three identical loops, full height and pressure - has been constructed with the main objectives of contributing to the verification of the CATHARE calculation code of accidents and to the validation of the physical bases of Emergency Operating Procedures.So far several tests have been (November 1988) carried out, among which natural circulation under various conditions (single phase, two-phase, symmetric or asymmetric conditions, different core power, variable steam generator liquid level), 2 inches cold leg break, steam bubble formation and collapse in the upper head. Summarized results of some of these tests are introduced and compared to a first CATHARE calculation as far as the 2″ cold leg break is concerned.     

压水堆核电厂蒸汽发生器传热管道破裂事故源项的计算分析

当压水堆核电厂发生事故后,带有放射性的核素会通过破损处释放到环境中,从而危害核电厂周边环境及相关人员的安全,因此对事故后释放到环境中的放射性源项分析,对于核电厂的辐射防护具有重要意义。本文根据事故发生的频率以及后果严重程度,选取蒸汽发生器传热管破裂事故(Steam Generator Tube Rupture,SGTR)进行分析。事故分为事故前碘尖峰释放和事故并发碘尖峰释放两种事故工况,建立事故后放射性核素迁移和扩散计算模型,同时使用先进压水堆AP1000参数进行计算验证,并重点关注惰性气体和挥发性核素碘在环境中的放射性活度。计算结果显示:使用文中计算模型计算的放射性源项与设计源项比较一致,在两种工况下,惰性气体的释放活度与设计源项吻合较好,但碘的释放活度有明显差别。     

Experimental verification of accident loads for PWR containments

The principal loads which a nuclear power reactor containment is designed to withstand are produced by internal fluid caused static and/or dynamic pressures. They can be generated during failure events which release mass and energy into the containment atmosphere. An overview of the events which can generate substantial internal loads is provided. Representative experimental programs initiated for the investigation of the relevant physical phenomena are described. Illustrative examples of measured data are presented and discussed.     

MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.     

Theoretical predictions of failure probabilities for PWR pressure vessels subjected to accident conditions

The Second Marshall Report (1982) presented a detailed analysis of the integrity of PWR pressure vessels. As part of that study theoretical calculations of failure probabilities were made. Since the publication of that Report modifications have been made to the theoretical model to extend the failure criterion into stable crack extension, to update the knowledge of the distribution of various parameters, to more accurately represent the stress intensities and crack shapes, and to consider a different representation of pre-service detection of defects.In this paper these modifications are summarised and the application of the model to the calculation of failure frequencies for the most severe accident conditions, the large loss-of-coolant accident and the steam break, is presented. The results indicate that the probability of vessel failure is one to two orders of magnitude lower than previously predicted.     

Severe accident mitigation concept of the EPR

For the EPR great emphasis is laid to gain further improvement in prevention of severe accidents. Nevertheless an additional level of safety with specific engineered safety features is introduced to cope with the potential consequences of a severe accident with an assumed core melt down. To prove and demonstrate the feasibility of these features supporting research and development are performed via close co-operations with national and European research centres.     

Study on severe accident mitigation measures for the development of PWR SAMG

In the development of the Severe Accident Management Guidelines （SAMG）, it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident （LOCA）, Steam Generator Tube Rupture （SGTR）, Station Blackout （SBO）, and Anticipated Transients without Scram （ATWS） in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management （SAM） such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.