多轴疲劳模型对比研究及其在核电结构热疲劳寿命预测中的应用

压水堆核电站余热排出系统冷热水混合区管道发现的热疲劳问题影响核反应堆的安全。本文通过一种采用单轴疲劳试验数据拟合疲劳寿命曲线，进而用于预测多轴疲劳寿命的分析方法，基于文献中的疲劳试验数据，对Dang Van模型、Matake模型和Fatemi-Socie模型进行了余热排出系统冷热水混合区管道材料304L不锈钢疲劳寿命预测结果的对比研究。基于余热排出系统冷热水混合区管道的三维简化有限元模型，分别应用Dang Van模型、Matake模型和Fatemi-Socie模型对管道热疲劳寿命进行了预测，并与试验结果进行了对比验证。研究结果表明，基于应变（含平均应力修正）的Fatemi-Socie模型比较适用于304L不锈钢的疲劳分析，其热疲劳寿命预测结果相对Dang Van模型、Matake模型较合理。     

余热排出系统管道热疲劳裂纹萌生方向的理论和数值研究

在压水堆核电站余热排出系统冷热水混合区管道发现了由热疲劳导致的不同方向的浅裂纹群,疲劳裂纹的萌生方向可根据临界面方向进行预测。本文对临界面方向进行了理论推导,得出了双轴疲劳载荷作用下临界面方向的解析解。基于C++语言开发了临界面方向的分析程序,同时采用C++开发的分析程序和有限元软件Code_Aster计算了临界面方向,并将计算结果与理论解析解进行了对比验证。临界面方向分析结果与余热排出系统管道中发现的热疲劳裂纹方向吻合。研究表明,余热排出系统管道焊接残余应力对热疲劳裂纹萌生方向具有决定性的作用。     

核辅助管道Farley-Tihange现象机理分析及对策研究

基于热疲劳的核辅助管道Farley-Tihange现象频繁出现在运行核电厂的余排热交换器下游冷、热水混合区,以及与主冷却剂管道直接相连的安全注入系统和余热排出系统管道焊缝或母材上,轻则引发超标缺陷、重则导致小LOCA（冷却剂丧失事故）。本文概括了Farley-Tihange现象典型位置和潜在风险,建立了失效原因故障树,分析了Farley-Tihange现象的热疲劳失效机理与模式,提出了包括运行监督、停堆无损监测和应急维修预案等在内的系统化应对策略。     

核电厂主管道材料低周疲劳寿命预测方法评价

采用总应变控制方法,对压水堆核电厂主管道国产材料Z3CN20.09M进行了室温与350℃温度下的低周疲劳试验研究,获得了材料的疲劳寿命演化规律。采用Manson-Coffin方程、单拉估算模型、拉伸滞后能寿命模型和三参数幂函数公式对该材料的低周疲劳数据进行了拟合。通过寿命预测结果比较发现,除单拉估算模型外,其他几种模型对350℃高温下疲劳寿命的预测结果分散性明显高于室温疲劳。在众多模型之中,单拉估算模型拟合效果较差且预测寿命偏于非保守,而室温下拉伸滞后能法预测精度相对较高,350℃下则采用三参数幂函数法获得的预测效果更好。     

中国实验快堆非能动余热排出系统中非能动爆破片疲劳性能试验

非能动余热排出系统是保证试验快堆安全工作的重要措施,爆破片是一种非能动超压泄放装置.本文对反复加载的疲劳载荷作用下的爆破片动作性能进行了实验研究,载荷应力为80%爆破片标定爆破压力,载荷循环次数超过1×105,疲劳试验前后的爆破片动作压力差别小于1.5%,说明余热排出系统中选用的爆破片耐疲劳性能优异,可以稳定保证非能动余热排出系统的可靠工作.     

先进压水堆核电厂主管道结构完整性的研究

以某先进压水堆核电厂主管道为例,对核安全一级管道的结构完整性进行分析评价,并对根据规范设计的管道设计裕量进行了分析。管道结构完整性评价内容包括依据规范对管道强度进行评价、采用解析法求解管道温度场进行热棘轮评价、采用简化雨流法对管道进行疲劳寿命评价。计算结果表明,主管道最小壁厚减少至55 mm能够满足标准规范要求,但安全裕度较小,其中主管道支管位置的疲劳和热棘轮评价结果裕量最小。     

核一级三通管热疲劳研究

钠冷快堆在启动和停止过程中会产生大的热应力,多次循环之后容易产生热疲劳损伤,特别是在三通管连接区域。本文将研究不同角度对三通管热疲劳性能的影响。通过ANSYS计算不同角度三通管道的热应力,确定三通管道的热疲劳寿命和疲劳损伤系数。给出了疲劳许用强度与三通管角度的函数关系。结果表明,随着角度的减小疲劳强度降低。此结果对核一级管道设计中选取三通管道的角度具有一定的参考价值。     

池式钠冷快堆事故余热排出系统一回路仿真研究

池式钠冷快堆事故余热排出系统采用了非能动工作原理,依靠液态钠及空气的自然对流排出堆芯余热。为研究事故工况下余热排出系统一回路的换热能力,基于FORTRAN语言,建立堆芯单通道及盒间流模型,采用全隐二阶迎风差分格式及改进的欧拉法离散求解,对事故余热排出系统一回路系统进行数值模拟,并对全厂断电事故进行仿真计算验证。结果表明:该程序能较好地反映事故余热排出系统瞬态变化过程,并可达到超实时仿真。     

快速疲劳分析方法在核电厂疲劳监测系统中的应用

疲劳监测系统通过对一回路易发生热疲劳关键管道和设备进行运行参数采集,采用快速疲劳分析方法对被监测管道和设备进行实时疲劳计算,从而获得真实疲劳损伤情况。该快速疲劳分析方法以格林函数法为基础,通过编制计算程序实现热应力和疲劳使用系数的快速计算。通过与有限元分析结果进行比较,验证了该快速疲劳分析方法具有高效、快速、准确的特点。      

核电厂仪表管道振动疲劳耐久性能改进研究

针对役前及初始运行期间核电厂出现较多的仪表管道振动疲劳断裂问题,设计了两套不同规格及焊接方式的、端部带有集中质量的悬臂管道试验件。对试验件开展了宽带随机耐久试验,分析了试样振动交变应力幅、频率响应特征及疲劳寿命,结果表明,通过采取增大管道外径和壁厚、改变焊接形式等措施,能够显著改进结构低频共振、试样振动疲劳寿命分散性较大等问题,显著提升管道结构的振动疲劳耐久性能。     

一回路相关死管段与热分层危害分析

为避免死管段与热分层危害，结合有关经验与核岛工艺系统设计特点，对某新型压水堆一回路各连接管逐一进行死管段与热分层危害分析。筛选出危害可能发生的管段后，对其中典型的热段连接余热导出管段应用计算流体力学软件CFX模拟分析，计算达收敛状态后可得出该管段热分层温度分布情况。另外，该管段下游两个隔离阀间封闭管段初始条件设定为充满工质，因受一回路影响而被加热升温，通过该封闭管段工质最终温度结果可判断是否出现死管段现象。最终计算数据显示热段连接余热导出管段总体上满足热分层验收准则，不过下游隔离阀间封闭管段有形成死管段的风险，但通过调整布置等措施可避免死管段危害。结果还显示出浮力循环流与一回路紊流冲击影响的流线特点。     

Transient behavior of the sodium–potassium alloy heat pipe in passive residual heat removal system of molten salt reactor

High temperature heat pipes, as highly-effective heat transfer elements, have been extensively employed in thermal management for their remarkable advantages in conductivity, isothermality and self-actuating. It is of significance to apply heat pipes to new concept passive residual heat removal system (PRHRS) of molten salt reactor (MSR). In this paper, the new concept PRHRS of MSR using sodium–potassium alloy (NaK) heat pipes is proposed in detail, and then the transient behavior of high temperature NaK heat pipe is numerically investigated using the Finite Element Method (FEM) in the case of MSR accident. The two-dimensional transient conduction model for the heat pipe wall and wick structure is coupled with the one-dimensional quasi-steady model for the vapor flow when vaporization and condensation occur at the liquid–vapor interface. The governing equations coupled with boundary conditions are solved by FORTRAN code to obtain the distributions of the temperature, velocity and pressure for the heat pipe transient operation. Numerical results indicated that high temperature NaK heat pipe had a good operating performance and removed the residual heat of fuel salt significantly for the accident of MSR.     

快堆中心测量柱高周热疲劳评定方法

流体温度振荡是快堆中的常见现象。快堆中心测量柱下封头处于流体温度振荡严重的位置,流体温度振荡会降低反应堆结构材料的寿命,威胁反应堆的安全。为了研究流体温度振荡对中心测量柱下封头寿命的影响,将中心测量柱下封头简化为平板模型,提出了一套热疲劳评定方法。利用Miner法则对结构材料的疲劳损伤因子进行计算,并针对中心测量柱下封头进行分析。本文计算结果符合Miner法则的要求,可为工程应用中的材料选取和厚度选择提供参考。     

Piping Cracks in JPDR, (I)

In JPDR, primary coolant leaked in August 1972 through a crack in the heat affected zone between nozzle safe end of pressure vessel and pipe. And other piping systems were then investigated, and cracks were detected in two systems. Then relations between stress and cracks in the pipes were studied. Stress was analyzed by elastic theory and examined by mockup fatigue test of pipe. Initiation of cracks are not caused by stress alone, indicating the existence of other factor involved.     

AP1000主给水管道断裂事故中PRHR系统冷却能力分析

使用机理性分析程序建立包括主冷却剂系统、专设安全设施及相关二回路管道的AP1000核电厂模型,对AP1000核电厂主给水管道断裂事故进程进行计算分析。着重分析了非能动余热排出（PRHR）系统在主给水管道断裂事故工况中的瞬态响应、热工水力行为及其冷却能力,并针对PRHR系统流道阻力特性的不确定性对冷却能力的影响进行分析。分析结果表明,在主给水管道断裂事故中,PRHR系统的热移出功率最终能够与堆芯的衰变功率相匹配,有能力带走衰变热,保证一回路系统最终处于安全停堆状态,不发生堆芯损伤,当PRHR系统阻力系数增加时,PRHR系统的流量和换热功率会降低,对PRHR系统冷却能力造成影响。     

Fatigue behavior of pipes containing multiple flaws in inner surface

In this work was studied the growth behavior of multiple cracks in the inner surface of pipes. The fatigue tests were performed using two kinds of test pipes, i.e., the straight pipes and bend pipes of AISI Type 304L stainless steel, having 320 mm in outer diameter and 35 mm in thickness approximately.The crack growth curves obtained by the fatigue tests were compared with the analytical curves of two kinds of crack growth prediction methods. One method is based on the ASME Boiler and Pressure Vessel Code, Sec. XI. Another method is based on the procedure in which the crack growth formula is applied to both the surface and thickness directions. The analytical crack growth curves predicted by the ASME Code are conservative for the test results of the straight and bend pipes. However the results of bend pipe test suggest that the procedure of the ASME Code may give an unconservative fatigue life under the certain condition.On the other hand, the test results of the straight pipes can be evaluated reasonably and those of the bend pipes can be evaluated conservatively by the latter method.     

Crack arrest in high cycle thermal fatigue crazing

Thermal crazing in high cycle thermal fatigue due to thermal fluctuation in residual heat removal (RHR) system of some nuclear power plants is explained by crack arrest in the depth due to a decreasing stress intensity factor. This is related to high frequencies of thermal loading. An attempt has been made through a parametric study to acquire some knowledge about the loading, knowing the crack depth. For this purpose, analytical as well as finite element simulations of crack propagation in 2D- and 3D-semi-elliptical cracks have been performed. In periodic loading, bounds for the number of cycles to fatigue life are proposed. Moreover, it is shown that in the absence of mean stress, fatigue damage in RHR may be produced in the macroscopic elastic-plastic regime. Finally, it is shown by FE simulations that for a semi-elliptical crack, a small error on stress intensity factor may result in significant error on crack length at high number of cycles, due to error accumulation cycle by cycle. Moreover in this paper is given the reason as to why shielding effect has not been taken into account in the study of crack arrest in RHR.