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.     

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.     

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。      

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.     

A numerical algorithm for analyzing piping response to thermal transients and seismic excitations

This paper describes a three-dimensional finite-element method for the structural response analyses of reactor piping systems subjected to thermal transients and seismic excitations. In the thermal-transient analysis, the algorithm appropriately accounts for stresses arising from the temperature dependence of the elastic material properties, the thermal expansion of the materials, and the change in the temperature-dependent yield surface. It includes a thermoviscoplastic constitutive equation that incorporates capabilities of treating thermal softening, failure, strain rate, creep, and stress ratcheting.In the piping seismic analysis, a time-history method is utilized to account for possible nonlinear phenomena as well as to reduce the solution conservatism obtained from response-spectrum analyses. Variation in dominant stress frequencies and translational frequencies are easily handled.The algorithm described herein has been integrated with the implicit time-integration scheme for treating hoop, flexural, axial, and torsion deformations induced by these accident loads. It is valid for large displacement and rotations, but small strains. Both geometrical and material nonlinearities are considered. To illustrate the methodology, several sample problems are given. Results show that this approach is extremely efficient and is suitable for long-duration analyses of reactor piping systems.     

热管堆堆芯基体结构高温力学行为分析

为研究热管堆堆芯基体结构高温下的热应力失效行为,以简化的多孔基体结构为研究对象,结合Megapower 5 MW（热功率）热管堆的设计参数,制定了正常工况和异常工况2种工况下的高温试验方案,其中异常工况考虑了单根热管失效。宏观检测结果显示基体结构未发生明显的变形与失效,结合数值分析方法获得了基体结构在2种工况条件下的温度分布和应力-应变响应,进一步说明了在试验条件下基体结构并不会发生静强度失效和塑性垮塌失效。本研究为明确热管堆堆芯基体结构的强度设计准则奠定了基础。     

Probability of crack-induced failure in BWR recirculation piping

The Lawrence Livermore National Laboratory (LLNL) has estimated the probability of double-ended guillotine break (DEGB) in the reactor coolant piping of Mark I boiling water reactor (BWR) plants. Two causes of pipe break are considered: crack growth at welded joints and the seismically-induced failure of component supports. For the former a probabilistic fracture mechanics model is used, for the latter a probabilistic support reliability model. This paper describes a probabilistic model developed to account for effects of intergranular stress corrosion cracking (IGSCC). The IGSCC model, based on experimental and field data compiled from several sources, correlates times to crack initiation and crack growth rates for Types 304 and 316NG stainless steel against material-specific ‘damage parameters’ which consilidate the separate effects of coolant environment (temperature, dissolved oxygen content, level of impurities), stress (including residual stress), and degree of sensitization. Application of this model to actual BWR recirculation piping shows that IGSCC clearly dominates the probability of failure in 304SS piping, mainly due to cracks that initiate within a few years after plant operation has begun. Replacing Type 304 piping with 316NG reduces failure probabilities by several orders of magnitude.     

Structural design of high temperature metallic components

In the high temperature engineering test reactor (HTTR), even at normal operation the service temperatures of class 1 metallic components reach temperatures above 900 °C when exposed to primary helium coolant of 950 °C. For these components, Hastelloy XR, which is the improved version of Hastelloy X, was developed and used for high temperature application.Some of the high temperature materials and their service temperatures, including Hastelloy XR, used for the class 1 and reactor internal metallic components of the HTTR are very well beyond the well-established Japanese elevated temperature structural design guideline. Moreover, at very high temperatures, where creep deformation is significant, the component design based on elastic analysis is impossible. Therefore, many research works on structural mechanics behavior were carried out to establish a high temperature structural design guideline and creep analysis methods. This paper reviews structural design of the high temperature components for the HTTR made of Hastelloy XR, 2 1/4Cr–1Mo steel, austenitic stainless steels SUS321TB and SUS316, and 1Cr–0.5Mo–V steel.     

示范快堆堆芯熔融物收集装置的安全分析

针对示范快堆堆芯熔融物收集装置的高温结构完整性问题,采用堆芯熔融物滞留在反应堆压力容器策略有效性评估方法（IVR-DOE10460）,建立了316H本构模型、多轴修正以及具体的分析评价方法。通过搜集与分析ASME规范和R66材料数据手册中316H钢相关的材料数据,确定了输入数据。在此基础上,利用有限元分析软件ABAQUS开展堆芯熔融物堆积形态下堆芯熔融物收集装置的应力应变分析,并基于时间分数法与延性耗竭法（应变分数法）对堆芯熔融物收集装置进行蠕变强度校核。有限元分析结果表明：堆芯熔融物收集装置在设计时间内可满足时间分数和应变分数小于1的蠕变强度考核要求,且满足竖直位移小于设计指标的功能性要求。堆芯熔融物收集装置在堆芯熔化严重事故后能保持结构的完整性。     

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。      

Creep failure time of 20% cold-worked type 316 stainless steel

The Robinson failure criterion is examined for its accuracy to predict the creep failure time of 20% cold-worked type 316 stainless steel under uniaxial and multiaxial stresses. Observed changes of slope in the log-log plots of maximum tensile stress versus isothermal rupture life are neglected, and large errors in prediction of creep failure time are found. A failure criterion that best explains the experimental behavior of 20% cold-worked type 316 stainless steel in uniaxial and multiaxial creep conditions is developed by incorporating the effective stress responsible for crack initiation and the maximum tensile stress responsible for crack propagation into the creep failure time equation.     

基于PSA的压水堆LBLOCA不确定性分析

为了结合确定论与概率论分析开展更加真实的核反应堆事故工况安全分析,提出了一种结合概率安全分析（PSA）和最佳估算加不确定性（BEPU）分析的方法,并以典型三环路压水堆冷管段双端断裂大破口失水事故（LBLOCA）的极限事故为对象,首先基于PSA开展了应急堆芯冷却系统的事故失效分析,而后结合BEPU分析评估了事件树中各事故序列的包壳峰值温度（PCT）分布及条件堆芯损坏概率（CCDP）,最终确定了压水堆在该事故工况中的堆芯损坏频率（CDF）。分析结果表明,压水堆在冷管段双端断裂工况中应急堆芯冷却系统能够保证反应堆的安全,且一列低压安注系统足以排出堆芯余热及保证反应堆安全。      

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。     

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.     

Development of flow-induced vibration evaluation methodology for large-diameter piping with elbow in Japan sodium-cooled fast reactor

This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in Japan sodium-cooled fast reactor, with particularly emphasis on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. The approach to the methodology development was defined: experiment-based methodology and simulation-based one as well as extrapolation logic to the reactor condition based on no dependency on Reynolds number in the high Reynolds number range from the experimental results. Experimental efforts have been made using 1/3-scale single-elbow test sections for the hot-leg piping as the main activity. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced pressure fluctuations onto the pipe though a slight deformation of flow separation was observed. Numerical results under different Reynolds number conditions appear in this paper using the unsteady Reynolds Averaged Navier Stokes equation approach, indicating its applicability to the hot-leg piping experiments.     

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.     

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。     

Mechanical modeling of creep, swelling and damage under irradiation for polycrystalline metals

A constitutive equation of creep, swelling and damage under irradiation for polycrystalline metals applicable to structural analyses in multiaxial state of stress is developed. After reviewing microscopic mechanisms of irradiation creep and swelling, the relevant theories proposed so far from the view point of metallurgical physics and their applicability are discussed first. Then a constitutive model is developed by assuming that creep under irradiation can be decomposed into irradiation-affected thermal creep and irradiation-induced creep. By taking account of the Stress-Induced Preferential Absorption (SIPA) mechanism, the irradiation-induced creep is represented by an isotropic tensor function of order one and zero with respect to stress, which is, at the same time, the function of neutron flux and neutron fluence. The volumetric part of the irradiation-induced creep is identified with swelling. The irradiation-affected thermal creep is described by modifying Kachanov-Rabotnov theory for stress-controlled creep and creep damage by incorporating the effect of irradiation. Finally irradiation creep and swelling of 20% cold-worked type 316 stainless steel at elevated temperature are predicted by the proposed constitutive equations, and the numerical results are compared with the corresponding experimental results.     

Results of reliability test program on light water reactor piping

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event.In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping.In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.