Numerical Simulations of Upper Plenum Thermal-Hydraulics of Monju Reactor Vessel Using High Resolution Mesh Models

In order to evaluate the effects of the FHs （flow holes） on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this model, it was mainly clear that in the 40% rated operational conditions, the shape of the FHs on the inner barrel did not change largely to the upper plenum thermal-hydraulics. The effect of the FHs on the honeycomb structure in the upper structure was also investigated in these calculations. The results indicated that the height of thermal stratification interface became lower than that evaluated from the test data.

Development of a new irradiation-embrittlement prediction model for reactor pressure-vessel steels

Predicting the transition-temperature shift(TTS)induced by neutron irradiation in reactor pressure-vessel(RPV)steels is important for the evaluation and extension of nuclear power-plant lifetimes.Current prediction models may fail to properly describe the embrittlement trend curves of Chinese domestic RPV steels with relatively low Cu content.Based on the screened surveillance data of Chinese domestic and similar international RPV steels,we have developed a new fluencedependent model for predicting the irradiation-embrittlement trend.The fast neutron fluence(E>1 MeV)exhibited the highest correlation coefficient with the measured TTS data;thus,it is a crucial parameter in the prediction model.The chemical composition has little relevance to the TTS residual calculated by the fluence-dependent model.The results show that the newly developed model with a simple power-law functional form of the neutron fluence is suitable for predicting the irradiation-embrittlement trend of Chinese domestic RPVs,regardless of the effect of the chemical composition.

Effect of weld microstructure on brittle fracture initiation in the thermallyaged boiling water reactor pressure vessel head weld metal

Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant.As-welded and reheated regions mainly consist of acicular and polygonal ferrite,respectively.Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions(0.5-2.5μm)at the brittle fracture primary initiation sites.High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch.The density,geometry,and chemical composition of the primary initiation inclusions were investigated.A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the debonded interfaces between the uncracked inclusions and weld metal matrix.Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance.

Effect of Pre-Deformation Enhanced Thermal Aging on Precipitation and Microhardness of a Reactor Pressure Vessel Steel

Microstructure evolution in neutron irradiated Reactor Pressure Vessel （RPV） steels was experimentally simulated through an improved degradation procedure in this study. The degradation procedure includes austenitizing at 1 150℃ and water quench, deformation 10% and 30% respectively, and then thermal aging at 500℃ for different period of time. The microstructure of the specimens was analyzed in details using transmission electron microscopy （TEM）. The micro-hardness test results showed that all the hardness curves of undeformed, 10% pre-deformed and 30% pre-deformed specimens have two micro-hardness peaks with the first peak value corresponding to different thermal aging time of 1 hour, 5 hours and 10 hours, respectively. It was revealed that the hardness curves were influenced by the precipitation of Cu-rich precipitates （CRPs） and carbides, deposition of martensite and work hardening.

Stress Analysis for Reactor Coolant Pump Nozzle of Nuclear Reactor Pressure Vessel

Integrated reactor structural design makes the pressure vessel itself and loads more complicated, so stress concentration makes strength failure easier at reactor coolant pump nozzle. The general purpose finite element program ANSYS/ WORKBENCH was used for 3D stress and fatigue analysis and the results of the evaluation are based on RCC-M criteria. The integrated reactor structural design is evaluated to demonstrate with applicable criteria and ANSYS/WORK- BENCH has better operability than ANSYS APDL on stress analysis of reactor pressure vessel.

Aging and Life Management System of Reactor Pressure Vessel

Reactor pressure vessel (RPV), the only key component that can not be replaced in nuclear power plants (NPPs), is the main barrier against the radioactive leakage. The lifetime of NPPs is dependent heavily on the life of RPV, and thus, the aging and life research on a RPV is a key factor in determining the life extension of NPPs. The purpose of this paper is to introduce an aging and life management system for an operating RPV which can be used as a reference of the lifetime extension. In order to realize the objective, an aging and life management system was developed. It is an comprehensive knowledge management system that integrates decentralized information and serves as a valuable data center. Based on the storage and management of RPV state information and operation data, this system provides real-time monitoring of important operating parameters, evaluation of irradiation embrittlement, and RPV aging assessment. Therefore, it is anticipated that the developed system can be used as an efficient tool for aging and life estimation of RPV.

堆内构件螺栓辐照促应力腐蚀可靠性评估方法

堆内构件连接螺栓所处工况环境恶劣,以辐照促进应力腐蚀开裂(IASCC)为代表的螺栓断裂失效事件时有发生,为了分析评估堆内构件螺栓的可靠性,提出了一种考虑螺栓IASCC失效的可靠性评估方法。首先,基于断裂力学理论中的应力强度因子准则,分别构建了SCC和IASCC失效的极限状态函数,并利用仿真分析方法求解函数中的最大等效应力。其次,考虑函数中各参数的不确定性,应用一次二阶矩法求解堆内构件连接螺栓的可靠度。最后,以堆内使用数量最多的M16型号螺栓为例进行计算分析。结果表明,断裂韧度与螺栓所受最大等效应力对螺栓可靠性的影响最大,裂纹扩展系数、裂纹扩展指数的影响次之,裂纹初始尺寸对可靠性的影响最小。此种评估方法可用于反应堆结构设计中各类连接螺栓的寿命预测及可靠性评估,也可以辅助确定反应堆结构安全检测周期和检测重点。

三代核电反应堆压力容器低合金钢焊缝性能对比分析

通过模拟国内三代核电压水堆反应堆压力容器环焊缝,分别采用国产焊接材料和进口焊接材料,在焊接工艺性能、焊缝金属化学成分、拉伸、冲击韧性、落锤和低周疲劳试验等方面进行了对比分析。结果表明,国产和进口焊接材料的焊接工艺性能和焊缝金属力学性能均相当;焊接试件不同厚度取样位置的焊缝金属-28.3℃冲击吸收能量平均值都大于90 J,有较大的设计裕量。该国产焊接材料已应用于国内某三代核电机组反应堆压力容器,为进一步推广应用提供技术支撑。

反应堆压力容器“set-on”集成化锻造结构多物理场耦合模拟研究

核反应堆压力容器(RPV)传统采用独立锻件通过焊缝连接的结构形式,由于焊缝为铸态金属组织,属失效分析中的薄弱环节,因此采用集成化锻造结构,确保承压焊缝质量,是提升RPV安全可靠性的重要方法。但是,集成化后的锻造结构存在几何异形、结构复杂、壁厚差异大等问题,其材料组织与性能调控难度大幅增加。针对上述问题,以核电RPV接管“set-on”结构形式为对象,基于结构导热微分及等温转变叠加理论,建立了锻造结构温度-组织-应力的多物理场耦合模型,以调质热处理的奥氏体化及淬火冷却两个阶段为对象,结合淬火冷却介质流-固耦合分析方法,获得锻造结构不同位置处材料组织与性能变化的定量规律,并明确了其关键影响因素。

国和一号关键核安全技术研发

基于当前各类能源技术基础和核电技术发展水平判断,核能是社会清洁低碳转型与可持续发展的重要支撑,高安全高可靠性的大型先进压水堆核电机组是未来30年内的主力机型。本文围绕大型先进压水堆核电站国家科技重大专项所面临的重大挑战,主要阐述了通过解决“高功率核燃料冷却难”“超高温熔融物滞留难”和“高温高压高放射性包容难”三大关键技术难题,来保证从设计上消除大规模放射性释放可能性或进一步降低核电批量化建设的核安全风险。

池式钠冷快堆熔融物堆内滞留初步分析研究

为防止堆芯熔毁后熔融物熔穿反应堆容器,造成大量放射性释放,三、四代反应堆设计中普遍考虑了熔融物滞留方案。池式钠冷快堆在主容器底部安装堆芯熔化收集器,对熔融物进行有效收集和长时冷却。利用中国原子能科学研究院自主开发的FRTAC程序,计算堆芯熔毁后主容器内的自然循环,分析熔融物长时冷却过程,研究钠冷快堆的熔融物堆内滞留方案。结果表明:熔融物掉落至堆芯熔化收集器上后,主容器内的自然循环可以有效冷却熔融物,并由事故余热排出系统将余热导出至大气环境中。

针对RPV钢磁巴克豪森噪声检测的传感器设计

辐照损伤是影响RPV使用寿命的重要因素,为了提升MBN信号检测仪器对RPV钢辐照损伤程度的评估性能,设计一种小型化MBN信号检测传感器;该传感器是由H型硅钢片磁轭、激励线圈、感应线圈和工型锰锌铁氧体构成,可以使检测仪器获得更为稳定的MBN信号,从而对RPV钢的辐照损伤进行有效评估;此外,利用电磁仿真软件ANSYS Maxwell对励磁模块进行了仿真与分析,获得试件的有效磁化区域,即长度8 mm(磁轭两脚之间的距离)、宽度8 mm(磁轭的厚度)、深度约为1 mm(试件的厚度)的立方体;为了降低激励信号对MBN信号接收的影响,分析了试件上方的磁感应强度分布,获得了MBN信号接收器应该位于磁轭两脚的中间位置;最后通过对RPV试样测量数据结果分析,确定了最佳激励信号的幅值为7 V,频率为6 Hz。

器内过滤对浆态床反应器操作特性的影响

浆态床反应器是低温费-托合成工艺普遍采用的反应器形式,常采用器内过滤系统连续分离反应生成的液体产品。通过实验方法从不同条件下过滤区域的漏气率与平均气含率的变化及气体反冲洗对床层操作稳定性的影响2个方面,研究了器内过滤对浆态床反应器操作特性的影响。结果表明:因为气体从过滤器处的泄漏,过滤区域平均气含率出现明显降低;过滤压差从8 kPa增至16 kPa,漏气率与气含率均随之增加,但平均气含率的平均增长值从0.0059降至0.0044;气-液体系的漏气率最大,平均气含率相对较低,范围为0.063~0.145;增加浆液颗粒质量分数到5%时,由于漏气率下降,平均气含率范围扩大至0.062~0.146。滤芯孔隙从80μm减小至30μm,气-液体系下漏气率降低幅度增大,平均气含率的平均增长值从0.0040增至0.0049;当加入质量分数为5%的颗粒后漏气率降低幅度开始下降,平均气含率的平均增长值也从0.0018降至0.0012。从过滤区域床层压差波动看,气体反冲洗对浆态床操作稳定性存在明显扰动,且滤芯孔隙越大,这种扰动效应越强烈。

反应堆压力容器无隔离层对接焊工艺研究

通过开展反应堆压力容器接管与安全端无隔离层对接焊模拟试验,以及焊接接头性能试验和无损检验,验证提出工艺的合理性及焊接接头可靠性,为反应堆压力容器接管与安全端焊接结构设计制造提供支撑。

反应堆压力容器压力-温度限值曲线分析方法研究

核电厂在反应堆启、停堆过程中必须将压力和温度控制在一定范围内,即压力-温度限值曲线(P-T曲线)所允许的范围内,以防止反应堆压力容器发生脆性开裂。本文论述了反应堆压力容器P-T曲线制定的防脆断设计的基础理论,总结了EJ/T 918-94规范和RCC-M 2007规范中关于P-T曲线的计算方法和分析流程,采用ANSYS APDL编程实现了P-T曲线的自动分析,并对RCC-M 2007规范中的P-T计算分析方法进行了完善,对计算结果进行了对比分析,研究表明EJ/T 918的计算结果过于保守。

池式快堆堆本体中液体晃动效应的研究进展

堆本体是核反应堆的核心组件,也是核反应堆进行核聚变或核裂变反应的关键部位。池式快堆堆本体内部的液态金属冷却具有大自由液面特征,其内部液体出现晃动可能会对核反应堆的安全性和稳定性造成不利影响,因此开展池式快堆堆本体中液体晃动效应的研究具有重要意义。首先,阐述了液体的晃动效应,并介绍了池式快堆堆本体的结构形式;其次,从理论研究、数值模拟和振动台试验三个方面对池式快堆堆本体的液体晃动研究进行了讨论分析。研究结果表明:单一的理论方法、数值模拟或振动台试验难以准确捕捉液体晃动对堆本体结构的实际影响,结合使用这三种方法能够获得更为准确和全面的研究效果;最后,对既有研究进行了梳理与总结,以期为池式快堆堆本体的进一步抗震研究提供新的思路和启发。

反应堆压力容器主螺栓孔的视频扫查装置

反应堆压力容器螺栓孔是压力容器顶盖设备的重要组成部分。由于每次换料大修均需通过主螺栓的旋出和旋入来开关反应堆压力容器顶盖,故主螺栓孔容易产生接触损伤。针对此问题,开发了一套螺栓孔内表面视频扫查装置,介绍了该装置的研制方案和系统构成,并对其进行了功能测试,对压力容器的螺栓孔进行检查,测试结果表明,该装置可以在压力容器法兰表面绕中心轴周向行走,运动准确,可靠性好,采集的视频图像清晰,畸变率低,满足使用需求。同时,该套视频扫查装置集成程度高,操作简单,兼具手动及自动模式,可显著降低现场工作人员承受的辐射剂量,提高在役检查的效率。

新型反应堆压力容器主螺栓螺纹副性能试验研究

以国内某新型号压水反应堆压力容器主螺栓螺纹副为研究对象,对其进行了各安装工况下的旋拧模拟试验和拉伸标定试验。结果表明:新型主螺栓螺纹副的各安装参数推荐值可为主螺栓安装偏心度应不大于0.5 mm,主螺栓旋转速度应不大于30 rad/min,主螺栓剩余平衡配重应不大于40 kg,主螺栓允许的最大旋转力矩可不超过90 N·m;新型主螺栓剩余拉伸量计算值与拉伸标定值几乎一致,新型主螺栓绝对拉伸载荷与剩余拉伸载荷比值约为1.22,在工程经验范围值内,证明了新型主螺栓螺纹副结构设计的合理性,且新堆型主螺栓在所有工况下拉伸过程中的最大允许绝对拉伸量为4.13 mm。

焊后热处理对核反应堆压力容器用16MND5钢组织和性能的影响

采用光学显微镜、透射电子显微镜,通过拉伸、冲击和落锤试验等方法研究了焊后热处理温度和保温时间对16MND5钢的组织和力学性能的影响。结果表明:经模拟焊后热处理后16MND5钢的室温和350℃抗拉强度和屈服强度都呈下降趋势,且下降幅度随焊后热处理温度的升高和保温时间的延长而增大;焊后热处理时间的延长导致钢的冲击性能小幅度下降,焊后热处理温度对冲击性能的影响并不明显;焊后热处理后的钢基体中弥散分布的细小第二相粒子数量减少、尺寸增大,导致钢的强度降低。

反应堆压力容器主螺栓表面缺陷的线激光测量与涡流检测对比

涡流检测是核电站反应堆压力容器主螺栓的主要检测方式,该方法主要用于对螺纹根部的缺陷进行表面检测。线激光测量技术是一种近年来新兴的非接触测量方式,通过高速激光扫描实现物体表面形貌的大面积、高分辨率测量,具有快速、非接触及高精度等特性。采用两种方法对反应堆压力容器主螺栓进行检测,试验结果表明,两种方法均可对其缺陷进行有效检测,线激光技术能精确到微米级,测量结果更加直观,且容易定量、定性,为主螺栓检测中的缺陷判定提供了可靠的数据支撑;两种检测方法各有优劣,实际检测时应取长补短,从而为反应堆压力容器主螺栓提供更加智能化、高效的检测方案。