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。     

严重事故下反应堆压力容器下封头耦合烧蚀传热分析

核电站严重事故发生后,反应堆压力容器（RPV）固壁在熔池作用下会发生烧蚀、减薄。开展RPV下封头耦合烧蚀传热分析对堆坑注水有效性论证和RPV剩余壁厚确认有重要的理论指导意义。本文以CPR1000反应堆压力容器为研究对象,在FLUENT 17.2平台下,基于动态网格方法和UDF二次开发,构建了综合考虑RPV固壁瞬态烧蚀与导热、RPV内壁热流密度再分布及RPV外壁过冷沸腾的全耦合计算模型,获取了9 000 s内的堆坑两相流场分布和RPV固壁烧蚀温度场,分析确定了最小剩余壁厚和发生位置。结果表明：使用动态网格捕捉壁面烧蚀的方法可行,本文全耦合计算模型在分析RPV固壁瞬态烧蚀过程方面有一定优势。     

Coupled Ablation and Heat Transfer Analysis of Lower Head for Reactor Pressure Vessel under Severe Accident Condition

After a reactor core melts accident, the solid wall of the reactor pressure vessel (RPV) will be inevitably eroded by the melting core which contains large density of heat flux. The analysis of the coupled ablation and heat transfer of the lower head for RPV is of great theoretical significance to the effectiveness demonstration of water injection in reactor pit and the confirmation of the residual wall thickness of RPV. In this work, numerical simulations were carried out based on the RPV model of CPR1000 using the CFD software FLUENT 17.2. Based on dynamic mesh model and user-defined function (UDF) redevelopment, a fully coupling calculation model considering the transient ablation and heat conduction of solid wall of RPV, the redistribution of heat flux density in RPV inner wall and the subcooled boiling of RPV outer wall was established. Both two-phase flow pattern in the reactor pit and temperature field of RPV solid wall ablation within 9 000 s were obtained and the minimum residual wall thickness and the occurrence location were determined by analysis. The results show that it is feasible to use dynamic mesh to capture wall ablation. The fully coupling calculation model has certain advantages in analyzing the transient ablation process of RPV under severe accident.     

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。     

Coupled thermal structural analysis of LWR vessel creep failure experiments

Considering the hypothetical core melt down scenario for a light water reactor (LWR) the failure mode of the reactor pressure vessel (RPV) has to be investigated to determine the loadings on the containment. The failure of reactor vessel retention (FOREVER)-experiments, currently underway, are simulating the thermal and pressure loadings on the lower head for a melt pool with internal heat sources. Due to the multi-axial creep deformation of the vessel with a non-uniform temperature field these experiments are an excellent source of data for validation of numerical creep models. Therefore, a finite element (FE) model has been developed based on a commercial multi-purpose code. Using the computational fluid dynamics (CFD) module the temperature field within the vessel wall is evaluated. The transient structural mechanical calculations are performed using a new numerical approach, which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a three-dimensional array is developed where the creep strain rate is evaluated according to the values of the actual total strain, temperature and equivalent stress. Care has to be exercised performing post-test calculations particularly in the comparisons of the measured data and the numerical results. Considering the experiment FOREVER-C2, for example, the recorded creep process appears to be tertiary, if a constant temperature field is assumed. But, small temperature increase during the creep deformation stage could also explain the observed creep behavior. Such considerations provide insight and better predictive capability for the vessel creep behavior during prototypic severe accident scenarios.     

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。      

熔融物堆内滞留条件下RPV长期结构完整性分析

熔融物反应堆压力容器(RPV)内滞留(IVR)是三代核电厂重要的严重事故缓解措施，而防止RPV的热工失效和结构失效是实现IVR的前提。本文建立考虑内壁面熔蚀的RPV有限元模型，在温度场分析的基础上，开展蠕变计算，得到不同时刻下的应力应变响应，通过选取典型评定路径并利用基于Larson-Miller参数的累积损伤理论进行蠕变损伤计算及评价。分析结果表明：在考虑一定内压的IVR条件下，RPV不会发生蠕变断裂，长期结构完整性可保证。本文的研究方法可为后续核电厂RPV在IVR条件下的结构完整性分析提供参考。     

熔融物堆内滞留条件下压力容器变形

熔融物堆内滞留(In-Vessel Retention,IVR)已经成为第三代反应堆一项关键的严重事故缓解策略,而压力容器外部冷却(External Reactor Vessel Cooling,ERVC)技术则是保证IVR得以成功实施的关键。当发生堆芯熔化时,高温熔融物对压力容器(Reactor Pressure Vessel,RPV)下封头的热冲击会导致RPV壁面和由其构成的外部冷却通道的形状发生变化,使局部传热恶化,进而造成IVR的失效。因此,有必要对IVR条件下RPV壁面的变形进行研究。本文利用有限元软件ANSYS对RPV进行了几何建模、温度场分析和力学场分析。结果表明,在RPV外部实现冷却、内部实现泄压的前提下,壁面变形为13.85-18.75 mm。在1 MPa内压的作用下,高温蠕变会使壁面变形随时间增大,但其增量有限。热膨胀是造成壁面变形的主要因素。     

国产反应堆压力容器用16MND5钢的蠕变损伤本构模型研究

为了获得反应堆压力容器（RPV）材料在高温下的蠕变行为,保证RPV在严重事故工况下的完整性,本研究对国产RPV用16MND5钢的高温蠕变性能进行了测试,获得了600~900℃下材料的蠕变性能,并基于应变强化的基本蠕变本构模型与基于延性耗竭理论的蠕变损伤模型,建立了适用于16MND5钢的蠕变损伤本构模型,给出了材料的蠕变损伤模型参数。结果表明,本文提出的蠕变损伤本构模型的有限元模拟数据与试验数据符合性较好,验证了此蠕变损伤模型的正确性。该方法可用于严重事故情况下RPV的蠕变损伤分析,为RPV的完整性分析提供支持。      

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.     

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.     

IVR-ERVC下压水堆压力容器下封头传热及应力/应变分析

以某1000?MW压水堆为例,利用二维极坐标热模型分析RPV壁面与双层堆芯熔池和外部冷却水堆腔之间的传热,计算下封头壁面瞬态二维温度场分布和烧蚀情况,同时通过有限元分析程序计算下封头壁面的各瞬态温度场和烧蚀引起的热应力/应变情况,分析压水堆RPV下封头在压力容器内熔融物滞留-压力容器外冷却（IVR-ERVC）下的结构完整性。计算结果表明:①芯熔融坍塌后200?s下封头壁面开始熔融,最薄厚度直线下降;3000?s后熔融区沿下封头内壁呈一片柳叶形状分布;②下封头内表面的吸热热流大于外表面的散热热流,在两层熔池界面处内外表面热流密度达到最大值;③RPV下封头热应力在0~400?s时集中于下封头内壁面;在400 s后,下封头内壁面热应力逐渐减小,形变量逐渐增大,下封头完整性可以得到保证;④2000?s以后,RPV下封头烧蚀损伤处内外壁面均产生应力集中,下封头烧蚀处内外壁应力值均大于许用应力,在2000?s后有可能发生断裂,在烧蚀损伤边缘处可能出现破口。      

承压热冲击下DVI接管动态应力特性研究

压力容器直接注入（DVI）接管在热冲击下的动态应力特性对于反应堆压力容器（RPV）结构完整性评估具有重要意义。建立了含DVI接管的RPV压力壳热流固耦合数值计算模型,并进行了验证分析;然后研究了蓄压安注箱（ACC）和堆芯补水箱（CMT）安注时RPV筒体和DVI接管热工水力特性;最后分析了热冲击下RPV筒体和DVI接管连接高应力区的温度分布、等效应力和等效塑性应变分布特性。研究结果表明,ACC安注阶段RPV筒体和DVI接管连接区存在较大的温度梯度和等效应力,且发生了局部塑性变形。若发生承压热冲击事件,应控制好DVI接管连接区温差,确保反应堆压力容器的结构完整性。本文开发的热冲击下热流固耦合数值计算模型和计算方法可用于核岛内DVI接管与RPV筒体的安全性评价,也可用于类似承压结构在热冲击下的动态应力特性分析。      

Simulation of scaled vessel failure experiments and investigation of a possible vessel support against failure

Scaled coupled melt pool convection and vessel creep failure experiments are being performed in the FOREVER program at the Royal Institute of Technology, Stockholm. These experiments are simulating the lower head of a pressurized reactor vessel under the thermal load of a melt pool with internal heat sources and a specified internal pressure. Due to the multi-axial creep deformation of the three-dimensional vessel with a prototypic non-uniform temperature field these experiments offer an excellent opportunity to validate numerical creep models. A Finite Element Model is developed and using the Computational Fluid Dynamic module, the melt pool convection is simulated and the temperature field within the vessel wall is evaluated. The transient structural mechanical calculations are then performed applying a new creep modeling procedure. Additionally, the material damage is evaluated considering the creep deformation as well as the prompt plastic deformation.After post-test calculations for the FOREVER-C2 experiment, pre-test calculations for the forthcoming experiments are performed. Taking into account both—experimental and numerical results—gives a good opportunity to improve the simulation and understanding of real accident scenarios. After analyzing the results of the calculations, it seems to be advantageous to provide a vessel support, which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it may be advantageous to install a passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.     

Fracture mechanical evaluation of an in-vessel melt retention scenario

This paper presents methods to compute J-integral values for cracks in two- and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.     

Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall.     

稳压器波动管蠕变破裂失效尺寸敏感性分析

以压水堆严重事故最佳估算程序为计算工具,研究了严重事故中稳压器波动管不同失效尺寸对严重事故进程和结果的影响。计算分析表明,稳压器波动管失效尺寸设为当量直径15cm左右的破口时可获得一个相对保守的计算结果,失效尺寸在12cm以下或18cm以上时,其计算结果没有15cm情况下的严重。研究结果可为深入研究压水堆核电厂严重事故现象提供参考。     

Bottom nozzle failure mechanism of water reactor pressure vessel under severe accident conditions

Most of past studies devoted to the creep rupture of a nuclear reactor pressure vessel (RPV) lower head under severe accident conditions, have focused on global deformation and rupture modes. Limited efforts were made on local failure modes associated with penetration nozzles as a part of TMI-2 vessel investigation project (TMI-2 VIP) in 1990s. However, it was based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failure has been investigated using data and nozzle materials from Sandia National Laboratory's lower head failure experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic–viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIP. It is concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure. From the FEA for the nozzle weld attached in RPV, it is shown that nozzle welds failure occur by displacement controlled fracture of nozzle hole not by load controlled fracture of internal pressure. Considering these characteristics of nozzle weld failure, new concept of nozzle failure time prediction is proposed.     

A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel – I: Basic concepts and model descriptions for the Fukushima Daiichi Nuclear Power Plant

This paper presents a simple approach for estimating the structure temperatures including the uncovered reactor core inside the reactor pressure vessel (RPV) and the release rates of fission products deposited in the RPV to the reactor building (R/B) at a certain time after the occurrence of a severe accident at a nuclear power plant (NPP). First, basic concepts are presented and then, a simplified steady-state heat balance model is proposed for estimating the temperatures of the uncovered reactor core and the upper structure in the RPV as well as the temperature of the RPV wall. In addition, models for estimating the revaporization rate of cesium hydroxide (CsOH) in the RPV and the leak rate of CsOH to the R/B via the drywell are also presented. The proposed approach is anticipated to be applicable to the damaged Units 1–3 of the Fukushima Daiichi NPP.     

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。