严重事故下封头失效机理分析

严重事故下堆芯熔融物坍塌到下封头,可能造成压力容器失效。本文针对造成压力容器失效的五个机制,运用一体化严重事故分析程序,分析全场断电分别叠加破口失水、主蒸汽输送管线破裂和蒸汽发生器传热管破裂事故对下封头完整性的影响。研究结果表明,三类事故均造成压力容器失效,全场断电叠加中破口失水事故由于破口位于热管段,距离稳压器和压力容器较近,事故响应更快,比全场断电分别叠加蒸汽发生器传热管破裂和主蒸汽输送管线破裂提前失效约20 000 s;全场断电叠加中破口失水事故中作用于贯穿件上的压力载荷超出贯穿件及其焊缝所能承受的最大载荷之和使得贯穿件弹出造成下封头失效;全场断电分别叠加蒸汽发生器传热管破裂和主蒸汽输送管线破裂均是因高温熔融物对下封头节点的损伤份额大于1使得下封头蠕变破裂造成压力容器失效。     

压力容器下腔室熔池结构计算方法研究

根据堆芯熔融物向下封头迁移的不同路径,给出压力容器下腔室内熔池结构的计算方法,并用MASCA实验结果对该方法进行验证。以百万千瓦级核电厂为对象计算全厂断电(SBO)事故工况下的熔池结构,结果表明,熔融物从侧面迁移到下封头,最终形成的熔池结构为3层。本方法可为熔融物堆内滞留条件下压力容器下封头的完整性判断提供条件。     

DVI管线破裂始发严重事故的IVR分析

本文选取了直接注入管线破裂始发的严重事故,分析堆芯熔融物压力容器内保持（IVR）策略实施以后压力容器下腔室内堆芯碎片和压力容器下封头的响应、堆芯碎片与压力容器壁面的传热、压力容器外壁面与堆腔水之间的传热以及压力容器不同区域的热流密度。研究表明,该事故序列下未发生下封头蠕变失效,区域4有最早发生蠕变失效的可能性。     

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。      

熔融物堆芯冷却滞留特性研究

以模块式小型堆ACP100为分析对象,建立MELCOR程序严重事故分析模型,分析了堆芯衰变热依次经过吊篮、压力容器壁面然后进入堆腔注水系统(CIS)的传热行为。采用燃料棒失效模型评价燃料组件坍塌行为,并通过ANSYS程序蠕变断裂模型评价堆芯下板失效行为。分析结果表明,严重事故后堆芯中心燃料组件坍塌形成堆芯熔融池,堆芯周围燃料组件保持完整结构状态,堆芯下板支撑堆芯熔融池和未坍塌的燃料组件且未发生蠕变断裂失效;CIS冷却压力容器外壁面并导出堆芯衰变热,最终实现熔融物堆芯滞留,避免下封头内形成熔融池。     

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。     

高温熔融物与金属堆腔相互作用实验研究

为掌握船用反应堆严重事故工况下压力容器失效初期堆芯熔融物热冲击对金属堆腔的破坏效应,开展了堆芯熔融物与金属堆腔相互作用机理实验。根据相似准则设计缩比金属堆腔实验装置,利用已有高温熔融物实验平台制备2 700 ℃高温氧化锆熔融物,通过特制卸料机构将高温熔融物卸料到实验段,对热冲击下实验段温度和变形响应特性及主要影响因素进行了研究。实验结果表明,高温熔融物进入金属堆腔初期,热冲击导致的金属堆腔最高温度为601 ℃,最大塑性变形量为0.44 mm,高温熔融物未导致金属堆腔热失效及断裂失效,金属堆腔实验段能保持完整。由于船用反应堆金属堆腔材料、结构和外部冷却条件更有利于保持金属堆腔完整性,基于实验结果推断,严重事故下压力容器下封头失效初期热冲击导致金属堆腔失效的风险较低。     

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。     

高温熔融物与金属堆腔相互作用实验研究

为掌握船用反应堆严重事故工况下压力容器失效初期堆芯熔融物热冲击对金属堆腔的破坏效应,开展了堆芯熔融物与金属堆腔相互作用机理实验。根据相似准则设计缩比金属堆腔实验装置,利用已有高温熔融物实验平台制备2 700℃高温氧化锆熔融物,通过特制卸料机构将高温熔融物卸料到实验段,对热冲击下实验段温度和变形响应特性及主要影响因素进行了研究。实验结果表明,高温熔融物进入金属堆腔初期,热冲击导致的金属堆腔最高温度为601℃,最大塑性变形量为0.44 mm,高温熔融物未导致金属堆腔热失效及断裂失效,金属堆腔实验段能保持完整。由于船用反应堆金属堆腔材料、结构和外部冷却条件更有利于保持金属堆腔完整性,基于实验结果推断,严重事故下压力容器下封头失效初期热冲击导致金属堆腔失效的风险较低。     

严重事故条件下压力容器完整性评价的研究进展

堆芯熔融物堆内滞留(In-Vessel Retention,IVR)是以AP1000为代表的第三代轻水反应堆严重事故管理的重要策略之一,也是严重事故条件下保证压力容器完整性(Reactor Vessel Integrity,RVI)的典型方法之一.该文综述了国外在严重事故条件下压力容器完整性试验研究和理论分析的现状,总...     

Sensitivity study of melt behavior of Fukushima Daiichi unit 1 type accident with MELCOR code and MPS method

Sensitivity calculation on melt behavior and lower head response at Fukushima Daiichi unit 1 reactor was performed with methods for estimation of leakages and consequences of releases (MELCOR) 2.1 and moving particle semi-implicit (MPS) method. Four sensitivity cases were calculated, considering safety relief valve (SRV) seizure, penetrations and debris porosity. The results indicated that the lower head failed due to creep rupture, not considering penetrations; otherwise it would have failed due to penetration tube rupture and ejection at an earlier time, resulting in part of debris dropping into the cavity of the drywell. The temperature of residual debris in pressure vessel kept low, and the vessel wall did not suffer creep failure up to 15 hours after reactor scram from which moment the water injection became available. Another aspect was that reactor pressure vessel (RPV) depressurization postponed the lower head creep failure time, and the low debris porosity brought forward the penetration rupture time. Either lower head creep failure or penetration rupture and ejection occurred in the central part of the pressure vessel. In MPS calculation, a slice of debris bed together with lower head, including an instrument guide tube, was chosen as the computational domain. Detailed temperature profiles in debris bed, penetration and vessel wall were obtained. The penetration rupture time calculated by MPS was earlier than the MELCOR result, while the vessel wall creep failure time was later.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

熔融物堆芯冷却滞留特性研究

The severe accident analysis model of the small modular reactor ACP100 is built using MELCOR code, and the core heat removed process through the barrel and wall of reactor pressure vessel (RPV) is analyzed by the cavity injection system (CIS). The collapse behavior of the fuel assemblies is estimated by the fuel rod degradation model, and the failure behavior of the lower core plate is estimated by ANSYS program. The results show that the fuel assemblies in the core center melt and collapse to form the core melting pool, while the structure of the fuel assemblies surrounding the core melting pool remains intact, and the core lower plate supports the core melting pool and un-collapsed fuel assemblies all the time, and no creep rupture phenomenon occurs; the core heat can be removed by CIS and the debris in-vessel retention successfully avoids the formation of molten pool in the lower head.     

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.     

Investigation of ALPHA experiment by severe accident analysis code SAMPSON

The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al₂O₃) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the code to simulate the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined where the results show its necessity in order to correctly reproduce the experimental trends.     

Risk Assessment of Creep Failure for RPV under Severe Accident Condition Based on Coupled CFD-FEM Method

After a reactor core melt accident, creep failure may occur in the residual solid wall of the reactor pressure vessel (RPV) under the influence of high temperature difference, internal pressure and the weight of the molten pool. In this work, the CPR1000 RPV was used as a research object. The ablation temperature field of the lower head of RPV was solved through the secondary development of the FLUENT software. And then, a CFD-FEM coupling analysis was carried out based on ANSYS Workbench software. The equivalent stress, the equivalent plastic strain and the equivalent creep strain of the RPV within 72 h under severe accident after the wall ablation and temperature field distribution formed stably were calculated. The risk of creep failure of the RPV was evaluated. The results show that when the reactor pit water injection measure puts into operation, the residual solid wall of the RPV will not experience creep failure and plastic deformation failure within 72 h, and besides, the pressure relief can significantly increase the safety margin of the structural integrity of the RPV.     

基于耦合CFD-FEM方法的严重事故下RPV蠕变失效风险评估

核电站严重事故发生后,反应堆压力容器（RPV）的剩余固壁在高温差、内压、熔池重量等的作用下可能发生蠕变失效。本文以CPR1000 RPV为研究对象,基于FLUENT软件二次开发求解反应堆压力容器下封头烧蚀温度场,然后基于ANSYS Workbench开展耦合CFD-FEM力学分析,求解严重事故下RPV烧蚀温度场稳定后72 h内的等效应力、等效塑性应变和等效蠕变应变,并评估了RPV的蠕变失效风险。结果表明：当堆坑注水等措施投运后,RPV剩余固壁在72 h内不会发生蠕变失效和塑性变形失效,有效卸压可明显提升RPV结构完整性的安全裕度。     

Global rupture of a nuclear reactor lower head with elevated temperature due to severe accidents

This paper is concerned with the global rupture of a reactor pressure vessel (RPV) with elevated temperature due to severe accidents in order to check if the RPV wall can retain the high-elevated pressure. The global rupture of an RPV is simulated by finite element limit analysis for the collapse load and mode to secure the safety criteria of a nuclear reactor under severe accident conditions. Finite element limit analysis is a systematic tool dealing with upper bounding and minimization technique to calculate the collapse load and mode. The finite element code (CALF, computer analysis of lower head failure) developed provides the temperature elevation in the lower head of a nuclear reactor under severe accident conditions as well as the collapse load and mode. The thermal analysis has to deal with heat transfer from the debris pool to the RPV wall and the top of the pool. The temperature distribution in such a system depends sensitively on the initial temperature of the debris pool and the thermal properties of a gap between the debris crust and the RPV wall. For accurate calculation, the thermal properties of a gap have to be determined in consideration of the gap size and conditions.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。