CPR 1000的IVR有效性评价中堆芯熔化及熔池形成过程分析

在发生堆芯熔化的严重事故后,通过容器外冷却将熔融物滞留在容器内(IVR)是一种重要的核电站严重事故缓解措施。本文通过选取与IVR有效性评价相关的严重事故序列,用一体化严重事故计算程序进行堆芯熔化过程计算及下封头中熔池的形成过程分析,得出下封头中分层熔池的结构和成分及其对金属层热聚集效应的影响。通过有、无容器外冷却模型的对比计算,评价CPR1000堆型的IVR的有效性。结果表明:在下封头熔池的金属层所在的高度上存在明显的热集中效应;而容器外冷却能保证压力容器的完整性。     

CRP1000的IVR有效性评价中堆芯熔化及熔池形成过程分析

在发生堆芯熔化的严重事故后,通过容器外冷却将熔融物滞留在容器内(IV)是一种重要的核电站严重事故缓解措施.本文通过选取与IVR有效性评价相关的严重事故序列,用一体化严重事故计算程序进行堆芯熔化过程计算及下封头中熔池的形成过程分析,得出下封头中分层熔池的结构和成分及其对金属层热聚集效应的影响.通过有、无容器外冷却模型的对比计算,评价CPR1000堆型的IVR的有效性.结果表明:在下封头熔池的金属层所在的高度上存在明显的热集中效应;而容器外冷却能保证压力容器的完整性.     

IVR熔池分层模型对压力容器安全裕量分析的影响

严重事故缓解策略熔融物堆内滞留（IVR）有效性评价方法中，关于压力容器下封头内的熔池结构是最具争议的问题。本工作对目前国际上采用的稳定熔池2层和3层结构，以及在熔池形成过程中可能形成的4层结构进行了比较研究，建立了这3种结构下的熔池分层传热模型，并分析了3种结构在不同反应堆功率水平下对压力容器有效性的影响。结果表明，压力容器安全裕量随反应堆功率的升高而减小，在4层熔池结构下发生压力容器熔穿失效的可能性最大。     

大功率先进压水堆IVR有效性评价中熔池换热研究

熔融物堆内滞留-压力容器外部冷却（IVR-ERVC）是一种重要的核电厂严重事故缓解措施。当前针对IVR有效性评价的方法主要是基于集总参数模型对下封头熔池换热进行分析。在大功率先进压水堆熔池集总参数法计算中,堆芯重量变大、压力容器尺寸增加会导致熔池自然对流换热中的瑞利数Ra ′增大。通过使用集总参数分析程序,对比研究熔池氧化层各换热模型的适用范围,计算大功率先进压水堆高瑞利数条件下稳态熔池的自然对流换热,模拟两层稳态熔池模型中压力容器外壁面的热流密度分布,对其进行选定严重事故序列下的IVR-ERVC有效性评价,并对堆内构件设计提出建议。     

Churchill-Chu关系式在极薄金属层内传热的适用性研究

在假设的堆芯融毁事故中,反应堆压力容器的下封头内可能会形成分层的熔池结构。底部重金属层的形成会导致熔池顶部的金属层高度逐渐降低,使得顶部金属层的侧壁热聚焦效应逐渐增强,压力容器有可能会失效。本文通过开展HELM?LR试验,针对薄金属层在低高度条件下的传热进行了研究。结果表明,Churchill?Chu径向传热关系式在低高度的条件下依然适用。Churchill?Chu关系式在低高径比且以水为工质的条件下,计算结果偏不保守。将Churchill?Chu关系式运用到反应堆的熔池结构案例中发现,随着氧化物层衰变功率的增大和薄金属层高度的降低,氧化物层的等温边界假设将不再适用;虽然薄金属层侧壁的热聚焦效应仍会随其高度的降低而逐渐增强,但增加速度变缓慢。     

增强IVR有效性的堆内注水策略研究

熔融物堆内滞留（IVR）是一项核电厂重要的严重事故管理措施,通过将熔融物滞留在压力容器内,以保证压力容器完整性,并防止某些可能危及安全壳完整性的堆外现象。对于高功率和熔池中金属量相对不足的反应堆,若下封头形成3层熔池结构,则其顶部薄金属层导致的聚焦效应可能对压力容器完整性带来更大的威胁。本文考虑通过破口倒灌及其他工程措施实现严重事故下熔池顶部水冷却,建立熔池传热模型,分析顶部注水的带热能力,建立事件树,分析顶部注水措施的成功概率及IVR的有效性。结果表明,通过压力容器内外同时水冷熔融物,能显著增强IVR措施的有效性。     

压水堆下封头多层熔池模型敏感性分析

下封头熔池模型是熔融物堆内滞留（IVR）有效性评价的重要模型,已在典型压水堆安全评价中得到广泛应用。传统的2层熔池模型和近年来提出的3层熔池模型,主要模拟熔池内熔融物的成分及热量的分配与传递过程,具有关系式复杂和强非线性的特点。为了为熔池分层模型以及严重事故缓解策略的优化提供帮助,采用中国核动力研究设计院自研的全局敏感性分析工具SALib和熔池分析软件CISER V2.0对4种熔池多层模型进行了敏感性分析,得到了主要输入参数对各模型关键结果参数的影响程度,敏感性分析结果反映了各熔池模型的典型特点。下封头半径对4种熔池分层模型均有显著的影响,Salay&Fichot模型与2层熔池模型中影响关键结果参数的输入参数基本相同,熔融物初始质量对Esmaili模型影响最大,熔融物密度对Seiler模型影响最大。      

基于ASTEC程序的反应堆严重事故下压力容器下封头传热分析

反应堆压力容器内熔融物滞留是先进反应堆设计严重事故缓解措施中的重要选项之一，在维持反应堆压力容器的完整性，包容堆芯熔融物方面具有重要作用。确保熔融物滞留有效性的关键是保证下封头内壁热负荷不超过下封头外壁面换热能力，而且在整个过程中不发生结构失效，即下封头剩余壁厚能够实现熔融物的承载。应用ASTEC程序，基于大型先进压水堆的设计，针对反应堆压力容器内熔融物滞留系统运行过程中冷却剂热工参数、下封头外壁面临界热流密度和最终下封头厚度进行计算分析，通过研究熔池对下封头的熔蚀和剩余厚度，判断下封头残留厚度对于熔融物的包容，评估系统的有效性。结果表明：在下封头较上部位置的部分区域内，换热较为剧烈，其中热流密度最大值出现在熔融物分两层的交界处，事故过程中下封头内壁将被熔融物金属层熔化，剩余厚度满足包容要求，但是最终剩余厚度十分有限。     

堆芯熔化严重事故下反应堆压力容器下封头高温蠕变分析

核电厂在发生堆芯熔化严重事故时，采用堆内熔融物滞留（IVR）策略将熔融物包容在反应堆压力容器（RPV）内是一项重要缓解措施。在IVR策略期间，RPV下封头在熔融物的极高温度载荷和力学载荷的共同作用下很有可能因过度蠕变变形而失效。因此，有必要对熔融物滞留条件下RPV下封头进行蠕变变形分析，以保证RPV结构完整性。该文在假定IVR条件下，采用有限元方法对RPV下封头进行热-结构耦合分析，通过计算得到容器壁的温度场和应力场，以及下封头的塑性和蠕变变形，并结合塑性和蠕变断裂判据对下封头进行失效分析。结果表明，考虑蠕变影响后，结构的变形将大大增加；严重事故下采取熔融物滞留策略期间，RPV下封头的主要失效模式为蠕变失效而非塑性失效；内压对蠕变变形量和蠕变失效时间有较大影响。该文为严重事故下RPV下封头的蠕变和失效研究提供了分析方法。      

三层熔融池结构情况下反应堆压力容器外水冷有效性分析

通过反应堆压力容器外水冷(ERVC)实现熔融物压力容器内滞留(IVR)是300 MW压水堆核电厂重要的严重事故管理特征。在过去IVR分析中通常对反应堆压力容器(RPV)下封头内两层熔融池结构进行分析,然而核电厂还可能出现一种底部为重金属层的3层熔融池结构,它可能对RPV完整性带来更大的威胁。本文根据建立的模型假设300 MW压水堆核电厂出现的该熔融池结构,并进行分析。结果表明,形成的底部重金属层不会威胁RPV完整性,但厚度变薄的顶部金属层失效裕度较小,可能威胁RPV完整性。     

Development and verification of molten corium–concrete interaction code

During a severe accident of Pressurized Water Reactor(PWR), the core materials was heated, melt located on the lower head of Reactor Pressure Vessel(RPV). With the temperature rise, the corium will melt through the lower head and discharge into the reactor cavity. Those corium will interact with the concrete and damage the integrity of the containment, so some coolability method should used to quench the corium. In order to investigate the progress of MCCI, a MCCI analysis code, that is MOCO, was developed. The MOCO includes the heat transfer behavior in axial and radial directions from the molten corium to the basemat and sidewall concrete, crust generation and growth, and coolability mechanisms reveal the concrete erosion and gas release, which are important for the interaction process. Cavity ablation depth, melt temperature, and gas release are the key parameters in the interaction research. The physical-chemistry reaction is also involved in MOCO code. In the present paper, the related MCCI experiment data were used to verify the models of the MOCO and the calculation results of MOCO were also compared with other MCCI analysis codes.     

Modelling of PWR lower head failure under severe accident loading using improved shells of revolution theory

In the study of severe pressurized water reactor accidents, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are usually investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exists. This may lead to an out of vessel steam explosion or to direct heating of the containment; both which have the potential to lead to early containment failure.Within the framework of the OECD Lower Head Failure (OLHF) programme, a simplified model based on the theory of shells of revolution under symmetrical loading was developed by IRSN. After successfully interpreting some other representative experiments on lower head failures, the model was recently integrated into the European integral severe accident computer ASTEC code. The model was also used to obtain the thermo-mechanical behaviour of a 900-MWe pressurized water reactor lower head, subjected to transient heat fluxes under severe accident conditions.The main objective of this paper is to present: (1) the full mathematical formulations used in the development of the model, including their matrices and integrals defined by analytical expressions; (2) the two creep laws implemented, one for the American steel SA533B1 and one for the French steel 16MND5; and (3) the various numerical interpretations of experiments using the simplified model. This paper can be considered as a theoretical manual to aid users of the simplified model during modelling of lower head failures under severe accident conditions. One of the applications presented in this paper concerns the determination of a diagram representing the vessel time to failure as a function of the pressure level and the heat flux intensity. This information has been used by IRSN in probabilistic safety assessment and severe accident management analyses.     

Analysis of safety margin of in-vessel retention for AP1000

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by most operating nuclear power plants and advanced light water reactors (ALWRs), AP600, AP1000, etc. External reactor vessel cooling (ERVC) is a novel severe accident management for IVR analysis. In present study, IVR analysis code in severe accident (IVRASA) has been developed to evaluate the safety margin of IVR in AP1000 with anticipative depressurization and reactor cavity flooding in severe accident. For, IVRASA, the point estimate procedure has been developed for modeling the steady-state endpoint of two core melt configurations: Configuration I and Configuration II. The results of benchmark calculations of AP600 by IVRASA were consistent with those of the UCSB and INEEL. Then, IVRASA is used to calculate the heat transfer process caused by two core melt configurations of AP1000. The results of calculations of Configuration I indicate that the heat flux remains below the critical heat flux (CHF), however, the sensitivity calculations show that the heat flux in the metallic layer could exceed the CHF because of the focusing effect due to the thin metallic layer. On the other hand, the results of calculations of Configuration II suggest that the thermal failure of the lower head at the bottom location is highly unlikely, but the heat flux in light metallic layer could be higher than that of base case due to the portion of metal partitioning into the lower head. This work also investigated the effect of the uncertainties of the CHF correlations on the analysis of IVR.     

船用堆严重事故下熔融物堆内滞留能力分析

堆芯熔化的严重事故，可能导致船用堆下封头失效、熔融物进入堆坑，危害人员及船体安全。本文采用严重事故一体化程序MAAP4，以船用堆全船断电事故为研究对象，针对低压安全注射系统投入时机、低压安全注射水流量，研究下腔室熔池形成后，投入低压安全注射系统对熔融物堆内滞留的作用。结果表明:在下腔室熔池形成后1576?s时，投入两台安全注射泵仍能有效阻止压力容器失效，实现熔融物堆内滞留；在下腔室熔池形成2646?s后，投入低压安全注射系统不能阻止压力容器失效。      

Consequences of material effects on in-vessel retention

In-vessel retention (IVR) consists in cooling the corium contained in the reactor vessel by natural convection and reactor cavity flooding. This strategy of severe accident management enables the corium to be kept inside the second confinement barrier: the reactor vessel. The general approach which is used to study IVR problems is a “bounding” approach which consists in assuming a specified corium stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium in the lower head. If there is no water in the vessel and if the corium pool is overlaid by a liquid steel layer, then the heat flux might focus on the vessel in front of the steel layer (“focusing effect”) and exceed the dry-out heat flux (CHF or DHF). One of the critical points of these studies is linked to the determination of the height of the molten steel layer that can stratify above the oxidic pool. The MASCA experiments have highlighted that part of molten steel may stratify under the oxidic corium which reduces the thickness of the steel layer on top of the pool. This behavior can be explained by chemical interaction between the oxide and metallic phases of the pool which confirms that these materials cannot be treated as inert species. Following these conclusions, a methodology which couples physicochemical effects and thermalhydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for given corium mass inventories. Attention focuses on the influence of parameters such as the ratio U/Zr and oxidation ratio of zirconia. For a 1000 MW PWR, approximately 10 t of steel stratify at the bottom of the vessel for 40% Zr oxidation, and 25 t for 30% Zr oxidation. This leads to a 25–50% increase of the mass of molten steel that is required for avoiding vessel melt-through.     

严重事故IVR下反应堆压力容器耦合传热数值模拟分析

堆芯熔化严重事故下保证反应堆压力容器（RPV）完整性非常重要,高温蠕变失效是堆芯熔化严重事故下反应堆压力容器的主要失效模式。在进行严重事故堆芯熔化物堆内包容（IVR）下RPV结构完整性分析中,RPV内外壁和沿高度方向的温度分布以及剩余壁厚是结构分析的重要输入。本文采用CFD分析方法对RPV堆内熔融物、RPV壁以及外部气液两相流动换热进行热-固-流耦合分析,获得耦合情况下的温度场、流场、各相份额分布以及RPV的剩余壁厚,为RPV在严重事故IVR下的结构完整性分析提供依据。     

Study of tearing behaviour of a PWR reactor pressure vessel lower head under severe accident loadings

In PWR severe accident scenarios, involving a relocation of corium (core melt) into the lower head, the possible failure mode of the reactor pressure vessel (RPV), the failure time, the failure location and the final size of the breach are regarded as key elements, since they play an important part in the ex-vessel phase of the accident.Both the LHF and OLHF experiments as well as the FOREVER experiments revealed that initiation of the failure is typically local. For the case of a uniform temperature distribution in the lower head, crack initiation occurs in the thinnest region and for the case of a non-uniform temperature distribution, it initiates at the highest temperature region. These experimental results can be modelled numerically (but more accurately with 3D finite element codes). The failure time predictions obtained using numerical modelling agree reasonably well with the experimental values.However, the final size of the failure is still an open issue. Analyses of both the LHF and OLHF experimental data (as well as of that from the FOREVER experiments) do not enable an assessment of the final size of the breach (in relation with the testing conditions and results).Indeed, the size of breach depends on the mode of crack propagation which is directly related to the metallurgical characteristics of the RPV steel. Small changes in the initial chemical composition of the vessel material can lead to different types of rupture behaviour at high temperatures. Different rupture behaviours were observed in the LHF and OLHF experiments using the SA533B1 steel. Similar observations were previously noticed during a CEA material characterization programme on the 16MND5 steel. To determine crack propagation and final failure size, 3D modelling would thus be needed with an adequate failure criterion taking into account the variability in behaviour of the RPV material at high temperatures.This paper presents an outline of the methodology being used in a current research programme of IRSN, in partnership with CEA and INSA Lyon. The aim is to model crack opening and crack propagation in French RPV lower head vessels under severe accidents conditions. This programme was initiated in 2003 and is made up of five main sections, namely an inventory of the different French PWR lower head materials, metallurgical investigations to better understand the cause of mechanical behaviour variability that is observed and related to material microstructure, Compact Tension (CT) testing of specimens to characterize the tear resistance of the material, validation of the modelling using experiments on tube specimens and the development of a new failure criterion for the 3D finite element models.     

ASTEC application to in-vessel corium retention

This paper summarizes the work done in the SARNET European Network of Excellence on Severe Accidents (6th Framework Programme of the European Commission) on the capability of the ASTEC code to simulate in-vessel corium retention (IVR). This code, jointly developed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS) for simulation of severe accidents, is now considered as the European reference simulation tool.First, the DIVA module of ASTEC code is briefly introduced. This module treats the core degradation and corium thermal behaviour, when relocated in the reactor lower head. Former ASTEC V1.2 version assumed a predefined stratified molten pool configuration with a metallic layer on the top of the volumetrically heated oxide pool. In order to reflect the results of the MASCA project, improved models that enable modelling of more general corium pool configurations were implemented by the CEA (France) into the DIVA module of the ASTEC V1.3 code.In parallel, the CEA was working on ASTEC modelling of the external reactor vessel cooling (ERVC). The capability of the ASTEC CESAR circuit thermal-hydraulics to simulate the ERVC was tested. The conclusions were that the CESAR module is capable of simulating this system although some numerical and physical instabilities can occur. Developments were then made on the coupling between both DIVA and CESAR modules in close collaboration with IRSN. In specific conditions, code oscillations remain and an analysis was made to reduce the numerical part of these oscillations. A comparison of CESAR results of the SULTAN experiments (CEA) showed an agreement on the pressure differences.The ASTEC V1.2 code version was applied to IVR simulation for VVER-440/V213 reactors assuming defined corium mass, composition and decay heat. The external cooling of reactor wall was simulated by applying imposed coolant temperature and heat transfer coefficient (HTC). The obtained results (pool temperatures, heat flux distribution, reactor wall ablation) were compared with available predictions of other codes. The agreement was correct, in particular on the shape and depth of ablation, as well as the maximum heat flux in case of a thick metallic layer, while ASTEC calculated a lower maximum heat flux for a thin metallic layer.