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。      

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.     

Finite element analyses of a BWR vessel and penetration under severe accident conditions

This study assesses the two-dimensional thermal response of a BWR vessel and drain line penetration to three types of debris bed; primarily metallic, primarily ceramic and metallic and ceramic layered, with sensitivity studies for the most severe case. Structural finite element analysis evaluates vessel elastic, plastic and creep response for two cases which bound the thermal challenge to the vessel.Thermal analysis results indicate that drain line failure does not occur for the case when metallic debris relocates to the lower head; structural analysis predicts that the vessel also remains intact for this case. In cases where ceramic debris relocates to the lower head, drain line temperatures peak near values where failure may occur within several minutes; whereas vessel failure is not predicted for 3.5 to 4.0 hours. Sensitivity studies indicate that large porosity debris or high heat removal rates from the vessel and drain line outer surfaces can preclude failure temperatures from occurring.     

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.     

Overview of NRC sponsored LWR lower head failure research program

The US Nuclear Regulatory Commission (US NRC) has sponsored a research program to investigate the mode and timing of vessel lower head failure. Major objectives of the program were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first for different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, the calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques were employed for analytical model verification and examination of more detailed phenomena. High-temperature creep and tensile data were obtained for predicting the vessel and penetration structural response. This paper summarizes major accomplishments and conclusions from research performed in the NRC sponsored lower head failure project.     

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。     

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.     

CORVIS. Investigation of light water reactor lower head failure modes

The project CORVIS (Corium Reactor Vessel Interaction Studies) is a research programme for the experimental and analytical investigation of a possible failure of the pressure vessel of a light water reactor during an assumed core melt accident. The applied methods of the experimental technique, the metallurgical investigations and the computational analysis are described. Two recent melting experiments are presented in detail. The first experiment was carried out on a steel plate 100 mm thick representing a pressure vessel lower head without penetrations. The second experiment was performed on a steel plate of the same thickness but representing a boiling water reactor lower head carrying a tube penetration in the form of a drain line. Reported are the phenomenology of these experiments, the results of the metallurgical post-test examinations and the computational reconstruction of the observed heat transfer and ablation phenomena in a stagnant and in a turbulent melt. Further, the results of three earlier tests, on a smaller scale, are presented.     

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.     

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.     

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.     

Assessment of reactor vessel integrity (ARVI)

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants.The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.     

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。     

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.     

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.     

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences.The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA.A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach.This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR,which is the representative case for high pressure sequences.MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident.In addition,a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.     

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。      

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.     

Results of examinations of pressure vessel samples and instrument nozzles from the TMI-2 lower head

15 prism-shaped steel samples were removed from the lower head of the damaged Three Mile Island Unit 2 (TMI-2) nuclear reactor pressure vessel to assess the effects of approximately 19 tonne of molten core debris that had relocated there during the 1979 loss-of-coolant accident. Metallographic examinations of the samples revealed that inside-surface temperatures of 800–1100°C were attained during the accident, in an elliptical ‘hot spot’ with dimensions of about 1 m × 0.8 m. Tensile, creep and Charpy V-notch specimens were cut from the samples to assess the mechanical properties of the lower head material at temperatures up to the peak accident temperature. These properties were used in a margin-to-failure analysis of the lower head. Examinations of instrument nozzles removed from the lower head region assisted in defining the relocation scenario of the molten core debris and showed that the lower head was largely protected from catastrophic failure by a solidified layer below the molten core debris that acted as a partial thermal insulator.     

Development of Debris Coolability Analysis Module in Severe Accident Analysis Code SAMPSON for IMPACT Project

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability.