严重事故下蠕变诱发RCS破裂的概率评估

严重事故下一回路管道可能会发生蠕变失效,若出现蠕变诱发的蒸汽发生器传热管破裂（SGTR）,则会导致安全壳旁路失效;若出现蠕变诱发热段或波动管的失效,则产生的破口将会使一回路迅速卸压。因此,评估严重事故下蠕变诱发反应堆冷却剂系统（RCS）破裂的可能性是开展严重事故分析、特别是二级概率安全分析（PSA）的重要基础。本工作基于蠕变失效模型,考虑传热管的缺陷,建立了评价蠕变诱发RCS破裂的确定论模型。在此基础上,运用拉丁超立方体抽样方法,考虑重要参数的不确定性,开发了严重事故下蠕变诱发RCS破裂的概率评估程序。随后对典型的事故序列进行了蠕变诱发RCS破裂的概率评估。结果表明,对于高压事故序列,存在一定的蠕变诱发SGTR概率,也存在较高的蠕变诱发热段或波动管失效概率。     

Effect of Microstructure on Failure Behavior of Light Water Reactor Coolant Piping under Severe Accident Conditions

In a severe accident of light water reactors, the reactor coolant system (RCS) piping might be subjected to thermal loads caused by the decay heat of the deposited fission products and the heat transfer from the hot gases, with an internal pressure in some accident sequences. Tests on the RCS piping failure were performed along with high temperature tensile and creep rupture tests including metallography to investigate the failure behavior. The prediction of the 0.2% proof stress by Arrhenius equation is in good agreement with the measured stress above 800°C for served RCS piping materials. The modified Norton's Law for the short term creep rupture model agrees with the experimental values between 800 and 1,150°C for type 316 stainless steel. The microstructural change was discussed with the effect of the very rapid formation and resolution of the precipitation on the strength at high temperature. The result of the piping failure tests which simulated the severe accident conditions, i.e., in short-term at high-temperature, could support the plastic limit load prediction of the flow stress model using the 0.2% proof stress.     

Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.     

Prediction of structural integrity of steam generator tubes under severe accident conditions

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800°C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models.     

管壁减薄对严重事故工况下传热管断裂的影响及缓解措施的有效性

严重事故的恶劣条件(反复的冷热交替及一、二回路之间的压差)可能导致蒸汽发生器(SG)传热管发生蠕变断裂。本文基于一级概率安全分析(PSA)的分析结果确定的典型事故序列,计算分析SG传热管壁减薄对严重事故工况下诱发蒸汽发生器传热管断裂(SGTR)的影响,给出严重事故缓解措施,例如一回路降压和给SG补水的有效性计算。     

小型压水堆严重事故下一回路承压管道蠕变预测分析模型开发

一回路承压管道蠕变是压水堆核电厂严重事故重要现象之一。针对小型压水堆,本文基于SCDAP/RELAP5程序开发了严重事故分析模型,利用实验拟合方法得到了一回路主管道(SA321)、自然循环式蒸汽发生器传热管(00Cr25Ni35Al Ti)两种材料蠕变预测分析模型,改进了SCDAP/RELAP5程序蠕变预测分析功能模块,并通过假想事故序列验证了SA321、00Cr25Ni35Al Ti蠕变预测分析模型的合理性。为后续开展小型压水堆严重事故下一回路承压管道蠕变规律研究提供基础参考。     

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。      

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.     

Methodology for developing channel disassembly criteria under severe accident conditions for PHWRs

This paper presents a methodology to develop a model for disassembly of the coolant channels in Pressurized Heavy Water Reactors under severe accident conditions. This model gives criteria to decide when under severe accident condition coolant channels will rupture due to deterioration in material properties at high temperatures and increase in load due to creep sag of channels above it and hence get disassembled. Presently available severe accident codes use simplistic and optimistic criteria based on a predefined temperature to predict failure of fuel channels and an explicit criterion for disassembly of the channel is not covered. The coolant channel disassembly model developed in this paper is based on modeling the sag and pile up of channels. A uniform temperature along the length of the channel is assumed. The disassembly of the channel is assumed when the total strain at any location exceeds the failure strain for a given temperature. A 3D failure surface which is a plot of time to failure, temperature of the calandria tube and load on the calandria tubes (on account of no of channels piled up) is developed. This failure surface can be used as an input to severe accident codes to predict the progress of the core disassembly. A set of failure surfaces is recommended to be used if metal–water reaction on the outer surface is to be accounted for loss in ductility due to metal water reaction. The temperature transient of the calandria tube for a severe accident obtained from system thermal hydraulic codes can be mapped onto the failure surface. The time at which the mapped transient crosses the failure surface gives the time at which the calandria tube is disassembled. This disassembly model is an engineered model which is much more realistic as compared to the current temperature based conservative model for predicting severe accident progression.     

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.     

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.     

Comparison of stress-based and strain-based creep failure criteria for severe accident analysis

We conducted a parametic analysis of stress-based and strain-based creep failure criteria to determine if there is a significant difference between the two criteria for SA533B vessel steel under severe accident conditions. Parametric variables include debris composition, system pressure, and creep strain histories derived from different testing programs and mathematically fit, with and without tertiary creep. Results indicate significant differences between the two criteria. Stress gradient plays an important role in determining which criterion will predict failure first. Creep failure was not very sensitive to different creep strain histories, except near the transition temperature of the vessel steel (900 K to 1000 K). Statistical analyses of creep failure data of four independent sources indicate that these data may be pooled, with a spline point at 1000 K. We found the Manson-Haferd parameter to have better failure predictive capability than the Larson-Miller parameter for the data studied.     

Semi-analytical modeling of a PWR lower head failure under severe accident conditions using an axisymetrical shell theory

The USNRC/SNL OLHF program was carried out within the framework of an OECD project. This program consisted of four one-fifth scale experiments of a reactor pressure vessel (RPV) lower head failure (LHF) under well controlled internal pressure and large throughwall temperature differentials; the objectives were to characterize the mode, timing and size of a possible PWR lower head failure in the event of a core meltdown accident. These experiments should also lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all ex-vessel events. A large quantity of escaping corium may lead to direct heating of the containment or ex-vessel steam explosion. These are important issues due to their potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, a 2D semi-analytical model has been developed and used to simulate these experiments. The aim of this effort is to develop a simplified but well predicting code that can be then implemented in European integral severe accident computer codes (ASTEC, ICARE/CATHARE). This paper presents the detailed mathematical formulation of this simplified method which is used to interpret the experimental results. The axi-symmetric shell theory under internal pressure proposed by Timoshenko has been utilised. The solution to the equilibrium equations is presented, with particular attention to the Rabotnov analytical formula. The radius and the polar angle of the deformed structure have been written as analytical expressions in order to take the large displacements and large strains into account using our mathematical formulation. The Norton type creep law and the Kachanov damage law have been used. Several failure criteria were used in the calculations and their effect on the numerical results is discussed. This 2D semi-analytical model gives very satisfactory results when compared, with the experimental and numerical results that were presented recently in the Benchmark calculations based on the first test of the OLHF program. The performance of this model is also illustrated by its capacity to accurately simulate the deformation of the lower head, including the variation of wall thickness.     

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.     

Effects of accident management strategy on the severe accident environmental conditions

This paper presents a methodology utilizing an accident management strategy in order to determine accident environmental conditions to be used as inputs to equipment survivability assessments. In the case that there is a well-established accident management strategy for a specific nuclear power plant (NPP), an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for the accident management strategy or appropriate actions. For this work, three different tools are introduced; probabilistic safety assessment (PSA) outcomes, major accident management strategy actions, and accident environmental stages (AESs). In order to quantitatively investigate an applicability of accident management strategy on equipment survivability, the accident simulation for most likely scenario in Korean standard nuclear power plants (KSNPs) is performed with the MAAP4 code. The accident management guideline (AMG) actions such as the reactor coolant system (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparison to actions from previous normal accident simulation, especially focusing on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages. This implies that plant-specific AMG actions need to be considered in order to determine accident environmental conditions in equipment survivability assessments.     

全厂断电引发的严重事故下蒸汽发生器传热管蠕变失效风险研究

全厂断电引发的严重事故若处置不当,可能发展为长期、高压的严重事故进程,此时堆芯冷却系统中的自然循环在导出部分堆芯余热的同时,也增加了蒸汽发生器（SG）传热管、稳压器波动管以及热管段出现蠕变失效的风险。本文基于两环路设计的秦山二期核电厂设计特点,结合蠕变失效风险模型,对全厂断电引发的严重事故后未能执行“严重事故管理导则中向蒸汽发生器注水（SAG-1）”时SG传热管的蠕变失效风险进行了研究,从而为全厂断电引发的严重事故的负面影响提供量化结果,为技术支持中心（TSC）最终决策提供参考依据。分析结果表明,全厂断电引发的严重事故后16 361 s可能出现蠕变失效;自事故后16 610 s,SG传热管出现蠕变失效的可能性均远低于稳压器波动管与热管段,秦山二期核电厂全厂断电引发的严重事故下因SG传热管蠕变失效而导致安全壳旁通的风险很小。     

Simple Method for Evaluating Mesh Size Effects in Control Rod Worth Analysis

It has been pointed out that the reactor coolant system piping could fail prior to the meltthrough of the reactor pressure vessel in a high pressure sequence of pressurized water reactor severe accidents. In order to apply to the evaluation of the piping failure which influences the subsequent accident progression, models for the strength of piping materials at high temperatures were examined. It was found that 0.2% proof stress and ultimate tensile strength above 1,073 K obtained from tensile tests was reproduced by a quadratic equation of the reciprocal absolute temperature. Short-term creep rupture time and minimum creep rate at high temperatures were well correlated by the modified Norton's Law as a function of stress and temperature, which implicitly expressed the effect of the precipitation and the resolution of precipitates on the creep strength. The modified Norton's Law gave better results than the conventional Larson-Miller method. Relating applied stress vs. minimum creep rate and tensile properties vs. applied strain rate obtained from the creep and tensile tests, a temperature range where the dynamic recrystallization significantly occurred was evaluated.     

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.     

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。