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。      

椭球形与球形下封头压力容器内熔融物滞留传热特性分析

严重事故下堆芯熔融物再分布于压力容器下封头，在衰变热作用下高温堆芯熔融物对压力容器壁面施加较大的热负荷，可能导致压力容器失效。针对压力容器内熔融物滞留下的传热过程，基于Fortran90语言开发了椭球形下封头压力容器内熔融物堆内滞留（IVR）分析程序IVRASA-ELLIP，计算具有椭球形下封头的压力容器在严重事故下稳态熔池的传热过程及IVR特性。利用IVRASA-ELLIP程序计算了VVER-1000压力容器内熔池的传热，分析具有椭球形下封头的压力容器各处的壁面热流密度、氧化物硬壳厚度和压力容器壁厚，并与运用IVRASA程序计算的AP1000稳态熔池传热结果进行对比分析。研究结果表明，在相同初始参数下椭球形下封头内的壁面热流密度较球形下封头内的小，与热流密度的变化趋势相对应，椭球形下封头内压力容器壁的消融量较球形下封头内的小，椭球形下封头内形成的氧化物硬壳厚度较球形下封头内的厚。     

IRIS反应堆严重事故下堆内自然循环及下封头失效分析

使用REALP5/SCDAP分析了IRIS堆汽轮机停机和部分失流事故导致的严重事故进程及缓解措施。分析结果表明IRIS堆内水装量大,使得堆芯较长时间处于淹没状态,事故发生后近7个小时堆芯开始裸露,10小时后堆芯开始损坏。对于不卸压不安注的情况,压力容器会完全干涸,堆芯和蒸汽发生器之间形成蒸汽自然循环流动,堆芯温度缓慢升高,低熔点的控制棒金属首先熔化落入下腔室并加热下封头,使得下封头底部区域发生蠕变断裂失效。在不卸压的情况下一个上充泵的安注流量就能够缓解事故。     

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。      

反应堆压力容器外水冷条件下贯穿件完整性分析

严重事敝下堆芯熔融物坍塌到反应堆压力容器(RPV)下封头时,可能造成贯穿件因高温熔融物热侵袭而失效,使压力容器丧失完整性,熔融物进入到反应堆堆腔中,导致熔融物堆内滞留(IVR)失效.在分析贯穿件脱落和熔融物流入贯穿件两种失效模式基础上,分别运用VTA程序和修正的整体凝固模型(MBF)计算贯穿件焊缝的熔化程度、热膨胀产生的摩擦力,估算贯穿件内熔融物流动的距离.结果表明,在成功实施反应堆压力容器外水冷(EVVC)措施条件下,300 MW压水堆核电厂压力容器的下封头不会因贯穿件失效而丧失完整性,堆芯熔融物小能通过贯穿件失效向堆腔迁移.     

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。     

ROAAM应用于ACP1000严重事故下实施IVR策略的有效性概率分析

基于堆芯熔融物与压力容器传热的机理分析模型,采用风险导向事故分析方法(ROAAM)分析压水堆在严重事故情况下通过冷却压力容器外部的手段来实施堆芯熔融物滞留在压力容器内(IVR)策略的有效性。以核电厂一级概率安全评价(PSA)分析结果为参考,计算ACP1000典型严重事故序列,分析影响熔融物传热的重要参数不确定性。概率分析结果表明:ACP1000发生假象的严重事故情况下,IVR策略有效性概率大于99%;由于熔融池顶部的金属层出现集热效应,下封头发生传热危险的主要位置出现在金属层。     

压力容器下封头大变形模型的开发及在FOREVER实验中的应用分析

反应堆发生严重事故后，将堆芯熔融物滞留在压力容器内的策略（In-vessel Retention,IVR）是作为缓解严重事故的一项重要措施，该策略已成功应用于AP1000、华龙一号和CAP1400等先进压水堆的严重事故管理中。在实施IVR策略时，下封头受到高温熔融物的热负荷会发生变形，下封头的变形改变堆腔的冷却流道，这会直接影响压力容器外部冷却的排热能力和IVR策略的成功实施，有必要对下封头变形展开研究和应用。针对ISAA(Integrated Severe Accident Analysis)程序LHTCM(Lower Head Thermal Creep Module)模型简化薄膜应力模型十分简单和缺乏计算变形模块的问题，本文从机理出发，基于Timoshenko板壳理论、Nortron蠕变定律和大变形塑性理论开发了机理模型—下封头大变形模型，并将该模型集成到一体化严重事故分析程序ISAA中对FOREVER-EC2实验进行应用，预测失效时间与实验的误差仅为1.9%，预测底部伸长量与实验测量值较为符合，破口位置与实验一致。分析结果表明该模型能准确预测在堆芯熔化严重事故中下封头所受应力、...     

全厂断电事故下封头熔池传热行为的研究

利用SCDAP/RELAP5系统程序对CPR1000核电厂进行了建模,并对全厂断电事故(SBO)的进程进行了模拟,分析了SBO中从堆芯开始裸露到完全裸露的熔化过程以及堆芯熔融物掉入下封头后下封头中熔池的传热行为。结果表明,熔融物在下封头形成一个混合层和重金属多孔介质层,且失效的位置在下封头侧部30°～40°位置(压力容器底部为0°)。     

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。     

Debris interactions in reactor vessel lower plena during a severe accident I. Predictive model

An integral predictive physico-numerical model has been developed to understand and interpret debris interactions in the reactor vessel plenum such as those which took place in the TMI-2 accident. The model represents the extent of debris jet disintegration by a jet-water entrainment model which can result in two types of debris configurations. One is particulated debris which eventually quenches in the water as a result of the entrainment process. The remainder of the debris penetrates to the bottom of the lower plenum and collects as a continuous layer. Each is treated as a separate region and has governing principles for its behavior. The potential for creating gap (contact) resistance and boiling heat removal is considered for heat transfer between the debris bed, the reactor vessel and steel structures and, most importantly, the vessel-to-crust gap water. The proposed in-vessel cooling mechanism due to material creep and water ingression into the expanding gap between the core debris and the vessel wall was found to explain the non-failure of the TMI-2 vessel in the course of the accident. The particulate debris bed is a mixture of metal and oxide, which is distributed as individual spherical particles of sizes determined at the time of entrainment. Energy is received from the continuum bed below by radiation and convection. The continuum debris bed is described by the crust behavior with the heat flux to the crust given by the natural convection correlations relating the Nusselt and Rayleigh numbers for the central region of debris. Using these governing principles, the rate laws for heat and mass transfer are formulated for each type of debris condition in the lower plenum. With the integration of the individual rates, the formation, growth and possible shrinkage of these regions are calculated. The potential reactor vessel breach is accounted for by considering the combined thermal and mechanical response of the vessel wall. The two-step failure model allows the vessel to fail at two different locations and at two different times.     

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.     

Debris interactions in reactor vessel lower plena during a severe accident II. Integral analysis

The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 °C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 Vessel Inspection Program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation.     

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。     

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.     

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.     

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.     

Core melt down and vessel failure: a coupled problem

Analyses of lower head failure have been performed for a variety of core slump scenarios that result from three contrasting reactor accident sequences in a PWR. The cases cover a range of thermalhydraulic conditions in the vessel and core debris characteristics. The results show lower head failure occurs at a time which depends on the internal thermal-hydraulic conditions and debris characteristics. Failure may be local or global and may be due to one or more of the following processes: creep; plasticity (including thermo-plasticity); and melt-through. At low to moderate pressure, creep damage accumulates over a wide area, leading to probable global failure. Local plastic deformation becomes increasingly important at higher pressures or following a pressure spike, with a possibility of local failure. Local melting can occur before failure if there is a large concentrated heat flux. A question of particular interest for future study is raised by the CORVIS experiments, namely that the deformation can cause a gap to open between the structure and debris crust and hence increase the thermal resistance. Modest estimates of the gap resistance show a significant delay in failure. A coupled treatment of the thermal and mechanical response is needed to assess the dynamic gap behaviour effectively.     

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。