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。     

CPR1000核电厂核岛主设备热棘轮现象试验评估方法研究

针对RCC-M评定准则对部分部件的热棘轮现象无法做出安全性评定的问题,综合考虑塑性变形和弹性变形的影响,详细分析CPR1000核电厂稳压器电加热喷雾接管的热棘轮现象。计算结果表明,RCC-M热棘轮评定准则过于保守,试验分析方法更精确,可提高核电建设的经济性。     

秦山核电二期工程冷却剂系统管件应力分析

用RCC-M B3650提供的简化分析方法及ANSYS程序,计算了秦山核电二期工程冷却剂系统接管嘴和焊缝的各类应力强度,分析了管道沿壁厚方向的瞬态温度场;用RCC-M提供的方法计算线性温差△T1和非线性温差△T2.结果表明,部分管件不满足RCC-M的方程(13)和热棘轮限制.     

应变幅对国产锻造奥氏体不锈钢环境疲劳寿命影响的试验研究

为验证模拟压水堆核电站冷却剂服役环境对国产锻造主管道用奥氏体不锈钢疲劳寿命的影响,采用高温高压循环水疲劳测试系统对从产品锻件取样加工后的标准试样进行了低周疲劳试验,分析了试验数据与美国机械工程师学会(American Society of Mechanical Engineers,ASME)规范平均/设计疲劳曲线的关系,获得了应变幅对奥氏体不锈钢环境疲劳寿命的影响规律,并初步评价了ASME规范设计疲劳曲线和环境疲劳修正系数的适合性。     

控制棒驱动机构耐压壳下部密封环应力与疲劳分析

采用有限元分析方法对某核电工程控制棒驱动机构耐压壳下部密封环的2种对接厚度进行了应力和疲劳分析对比,在疲劳分析中采用瞬态分组技术,同时参考RCC-M 2002规范对ANSYS程序中的弹塑性修正因子(Ke)进行解耦修正。结果表明,2种接头厚度的分析结果均满足RCC-M规范中的应力评定准则,其中,较薄密封环结构疲劳分析结果相对更安全,较厚密封环结构在其余工况相对更安全;在疲劳分析中对瞬态进行分组能明显降低使用系数和一次加二次应力之和幅值的保守性;在热和机械共同作用的一次加二次应力之和的幅值较高时,对Ke的修正能明显提高计算结果精度。     

针对热分层现象的稳压器波动管结构评价方法研究及分析程序应用开发

稳压器波动管热分层现象对核电厂安全运行具有潜在威胁。根据热分层发生机理,采用Fr数来判断热分层现象是否发生,研究热交换系数的计算方法,并将热分层引起的三维热应力解耦成一维总体应力和二维局部应力。根据RCC-M规范的要求,采用一维和二维组合的应力计算方法,将热分层产生的应力和其他载荷产生的应力叠加,进行结构完整性评定。配套本文提出的分析评价方法,对SYSTUS程序和ROCOCO程序进行应用开发。采用本文的分析评价方法和配套的分析程序,对秦山二期扩建工程稳压器波动管热分层进行分析评价,结果表明:稳压器波动管在热分层效应下结构完整性仍然满足RCC-M规范要求。     

反应堆冷却系统主管道疲劳暨最小壁厚分析方法研究

采用有限元法替代温度场差分方程计算温度瞬态在主管道壁厚方向上的温度分布,将温度计算结果与标准规范的计算公式相结合,从而求解各瞬态交变应力幅,以最终完成先进压水反应堆冷却剂主管道疲劳评定;通过疲劳求解的计算方法研究,提出最小壁厚的优化算法的迭代求解流程,可以依此通过编程最终实现疲劳评价和最小壁厚求解。     

压水堆-回路管道的铸造工艺及其国产化

工艺评定表明,1 000 Mw压水堆核电厂(CPR1000)原选用的主管道铸件Z3CN20-09M(法国牌号)不锈钢的化学成分符合RCC-M采购技术规范,但力学性能并不能完全满足压水堆核岛机械设备设计和建造规范(RCC-M)的要求.本文从金属学角度分析了Z3CN20-09M不锈钢抗蚀性特点和力学性能强化机理,确立了主管道铸件冶炼化学成份的内控标准,使CPR1000核电厂核岛主管道铸件(以下简称主管道铸件)的工艺评定在保持抗蚀性和可焊性特点前提下,各项力学性能指标均满足RCC-M标准,且有较大的裕度,离散度小,质量稳定,综合性能达到领先水平.     

核电厂核2级承压管道抗震设计规范对比分析

RCC-M、ASME(2007版)及GB 50267-97为目前核电厂设备、系统、部件设计所遵循的主要技术标准,3者对核电厂部件的分级基本相当,在核2级承压管道设计方面的规定内容相似但不完全相同.在地震输入方法上,GB 50267-97、ASME(2007版)及RCC-M基本相同,GB 50267-97中硬土场地的水平...     

化学和容积控制系统下泄热交换器抗震分析

采用有限元方法对辽宁红沿河核电厂一期工程设备中的化学和容积控制系统(RCV)下泄热交换器进行了抗震计算分析,载荷包括自重、压力、温度、接管载荷和地震.根据RCC-M和ASME规范对计算结果进行评定.结果表明,RCV下泄热交换器的设计满足规范要求.     

核安全设备疲劳分析方法与步骤

以某核电厂稳压器上封头为例,采用ANSYS有限元分析软件,考虑两个有代表性的温度和压力瞬态,同时考虑地震的影响,采用RCC-M规范,进行了疲劳分析,详细说明了核安全设备疲劳分析的方法和步骤。本文为核安全设备及管道系统的疲劳分析与评价提供了参考。     

Fatigue design of nuclear class 1 piping considering thermal stratification

In ASME B&PV Code, Section III, Subsection NB-3600, thermal stratification is not taken into account to determine the peak stress intensity range for fatigue design of nuclear class 1 piping. Therefore, the effects of several parameters such as boundary layer thickness, temperature difference, stratification length, wall thickness, inner diameter and material properties on peak temperature and peak stress intensity due to non-linear temperature distribution of thermal stratification in a pipe cross-section are studied through the numerical parametric study. The results of the parametric study are closely examined and consolidated to introduce an additional term into the equation of ASME so that the modified equation can be used to determine the peak stress intensity range due to all loads including thermal stratification.     

Fracture strength and behavior of carbon steel pipes with local wall thinning subjected to cyclic bending load

The power plant piping is designed to withstand seismic events using the design fatigue curve. However, the fatigue strength of a pipe with local wall thinning caused by erosion/corrosion is not clear. To evaluate the fatigue strength of pipes with local wall thinning, low cycle fatigue tests were conducted on 100A carbon steel pipes with local wall thinning. In load controlled tests on these pipes, ratcheting deformation was observed, and the fatigue strength became lower than that of cracked pipes. In displacement controlled tests, the fatigue strength of eroded pipes with 100 mm in eroded axial length, 0.5 in normalized eroded depth and 90° in eroded angle was almost equal to that given by the design fatigue curve in ASME B&PV Code Sec. III. To evaluate the local strain range in the maximum wall thinning area, the finite element analysis was conducted on the eroded pipes in the displacement controlled tests. It is concluded that the Mises strain range in the maximum wall thinning area and the low cycle fatigue curve can be used to conservatively estimate the low cycle fatigue life of an eroded pipe and the validity of estimated results can be confirmed experimentally.     

Normative issues in thermal fatigue design of nuclear components

The applicability of the presently existing nuclear standard codes (ASME and RCC-MR) in the thermal fatigue design of nuclear components is discussed with particular attention to the first wall of a fusion reactor machine. The possible extension of the results of the mechanical and thermo-mechanical fatigue tests, the sensitivity of the different numerical models in the calculation of the stress and strain history as well as the influence of the manufacturing technology on the thermal fatigue lifetime are tackled.     

Structural evaluation of a piping system subjected to thermal stratification

Piping systems in nuclear power plants are often designed for pressure, mechanical loads originating from deadweight and seismic events and operating thermal transients using the equations in the ASME Boiler and Pressure Vessel Code, Section III. In the last few decades a number of failures in piping have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. A simplified method has been developed in this work to estimate the stresses caused by the circumferential temperature distribution from thermal stratification. It has been proposed that the equation for the peak stress in the ASME Section III piping code include an additional term for thermal stratification.     

民用核安全设备活动中不符合(项)的识别、分类与控制

结合我国民用核安全设备活动管理现状,分析总结民用核安全设备活动许可证申请审查和监督检查中发现的不符合(项)识别、分类和控制中存在的问题,并参考Safety Series No.50-C/SG-Q,ASME NQA-1,RCC-M等国际通用标准规范的相关要求,提出民用核安全设备活动中不符合(项)的识别、分类和控制要求。     

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.