基于ASME和RSE-M规范的含平面缺陷奥氏体不锈钢核级管道剩余寿命预测方法的数值对比研究

核级管道在加工和安装环节可能存在不同的缺陷。此外,由于核电厂运行条件的影响,管道中可能存在少量缺陷,如裂缝。需要合理预测评估含缺陷管道的剩余寿命,以便安排更换方案,避免对核电厂的效率造成严重影响。本文根据ASME和RSE-M规范,在应力强度因子计算、裂纹扩展分析和裂纹稳定性评价等环节,通过数值对比研究了含有平面缺陷的奥氏体不锈钢核级管道的剩余寿命评估方法,为类似工作提供参考。      

两裂纹在核管道中应力强度因子的相互影响研究

随着核电站的老化,核管道安全评估变得更为重要。以往核管道的损伤容限评定通常只考虑单一裂纹问题,或只局限于研究一个主裂纹的问题,或主要集中于考虑两个裂纹满足何条件时,二者方可合并为一个裂纹的问题,对多裂纹情况下应力强度因子变化规律的研究较粗糙。在理论与方法方面,多裂纹问题一直是断裂力学不够完善的部分。本文以核管道的内表面裂纹为对象,采用有限元法分析核管道两个裂纹情况下的应力强度因子,得到其随裂纹间距、裂纹长度、裂纹深度变化的规律。又两个裂纹的现象为多裂纹现象中最基本的情况,即本研究工作可为存在多裂纹的核管道的安全评估提供方法依据。     

在役核压力管道模糊失效概率的计算方法

含缺陷核压力管道的失效概率实质上是一个模糊随机概率.本文应用模糊随机概率理论,同时考虑压力管道评定参数的随机性和失效模式判定的模糊性影响,提出了计算含缺陷核压力管道模糊失效概率的方法.与传统的压力管道可靠性分析方法相比,本方法考虑了模糊失效区域,评定的结果更为精确.     

残余应力作用下的不锈钢管道环向穿壁裂纹闭合效应研究

管道环向穿壁裂纹在不同载荷水平作用下的张开位移预测值是破前漏技术应用的关键核心参数。针对具有代表性几何尺寸的奥氏体不锈钢管道，采用数值分析和对比验证相结合的方法，基于工程中实际测得的材料性能曲线研究了典型焊接残余应力作用下穿壁裂纹临界闭合应力的变化规律。分析结果表明，目前的通用电气有限公司/美国电力研究院（GE/RPRI）方法和美国核管会技术报告NUREG/CR-6837修正方法均低估了由美国机械工程师协会（ASME）规范工作小组推荐的简化残余应力场所导致的管道环向穿壁裂纹闭合效应。此外，分析了环向穿壁裂纹闭合状态下管道的失效模式，在此基础上进一步讨论了裂纹闭合效应对破前漏技术应用的影响，为后续工程实践提供了可借鉴的技术观点。     

压力管道破前漏分析的一种简化方法

介绍了一种用于核反应堆管道和压力容器破前漏（ＬＢＢ）分析的简化方法，它主要以线弹性断裂力学的基础，将ＬＢＢ分析中应力强度因子，裂纹张开面积和泄漏率等计算以解析公式表达出来。这种方法使用方便，而且满足有关的国家标准和国际规范的要求，适用于一些应力分布和几何形状比较简单的管道和压力容器，或用于ＬＢＢ性质的近似估计。     

核级管道材料高温疲劳裂纹扩展行为实验研究

为研究核级管道材料在500℃以上的高温疲劳裂纹扩展性能,对管道母材、焊缝和热影响区材料进行了高温条件下的疲劳裂纹扩展速率试验,基于概率分析方法获得了考虑不同存活率的概率疲劳裂纹扩展曲线。研究结果表明,高温条件下,管道不同位置区域材料的疲劳裂纹扩展性能存在较为明显的差异,焊缝和热影响区的抗疲劳裂纹扩展能力明显优于母材。试验研究结果可用于核反应堆管道结构安全评估和断裂力学分析。     

压力管道“先漏后破”评定的准则研究

以弹塑性断裂力学为基础,分别建立了轴向表面裂纹和周向表面裂纹韧带的极限失稳压力Pu和穿透裂纹的起裂压力Pc的表达式,提出了一种压力管道的"先漏后破"缺陷评定准则:即若管道表面裂纹韧带的极限失稳压力Pu低于相应穿透裂纹的起裂压力Pc,则管道会泄漏失效;若管道表面裂纹韧带的极限失稳压力Pu等于或大于相应穿透裂纹的起裂压力Pc,则管道会爆破失效.该准则得到了一些文献提供的试验数据的验证.     

反应堆冷却剂系统辅助管道缺陷焊缝疲劳分析与评价

根据某核电厂反应堆冷却剂系统辅助管道核1级焊缝的在役检查结果和施工设计阶段应力分析结果,确定了疲劳分析与评价的典型缺陷焊缝.依据WRC502的实验结果和RCC-M规范,提出了用于疲劳分析的含热(微)裂纹效应的疲劳曲线.在此基础上,对机组运行5 a的瞬变统计次数与设计瞬态次数进行了对比研究,采用优化疲劳分析方法对典型缺陷焊缝进行了疲劳分析与评价.评价结果表明:辅助管道核1级焊缝在核电厂运行10 a内不会发生疲劳失效.     

含内表面裂纹核级管道的初始塑性失效分析

我国在役和新建的大部分核电厂在主管道上应用了破前漏技术,针对该技术ASME采用净截面屈服准则对完全塑性断裂进行缺陷评定,大量研究表明,净截面屈服准则高估了结构的承载能力。本文采用有限元方法模拟了含内表面裂纹的核级管道在内压作用下的变形过程,并利用裂纹前沿J积分随内压变化的曲线特征确定了含裂纹管道的初始塑性失效载荷。随后,将初始失效载荷的计算值与ASME规范定义的理论值相比较,结果表明理论解高估了结构的承载能力。最后,评价了ASME-BPVC-XI规范中A级使用限制对应的允许薄膜应力的适用性。     

失效评定图（FAD）技术在核管道缺陷评定规程编制中的应用

本文介绍了失效评定图（ＦＡＤ）技术的概念，用ＦＡＤ进行管道失效模式的判别，以及裂纹稳态撕裂扩展分析，进而预测承载能力的方法。对ＡＳＭＥ锅炉压力容器规范Ⅺ卷中核管道缺陷评定规程的技术难点以及在使用ＦＡＤ时所做的特殊处理及其背景作了分析和探讨。     

AP1000核电厂二回路主管道双端断裂流体喷射力计算分析

AP1000是先进的第三代压水堆核电厂,为确保核电厂在事故工况下的安全性,需对二回路主管道发生双端断裂的工况进行研究。本文采用RELAP5/MOD3.4软件对核电厂二回路突发主管道双端断裂的事故工况进行了数值模拟,计算得到断裂后管道破口处的喷放流量、压强、空泡份额及喷射力等物理参数的变化特性,并将计算结果与ANSI 58.2简化计算方法的结果进行了比较分析。结果表明,RELAP5/MOD3.4计算所得的喷射力小于简化计算方法所得结果。本文分析结果为进行AP1000核电厂的破裂管道甩击防护提供了基础。     

Precracked pipe under waterhammer action

We numerically simulate a full scale test in several computational steps with the finite element method and compare all calculated data with the experimental findings. First, we compute the deflection under static loading and the spectrum of eigenfrequencies of an integer piping, attached to a nuclear reactor pressure vessel (RPV). Then we consider a sudden pipe break at some distance from the vessel, immediately followed by an undamped closure of a check valve close to the break on the RPV side, and calculate the elastic and plastic transient dynamic response of the integer piping part between the RPV and the break. Finally, we consider a circumferential internal surface crack, fairly close to the vessel; after extensive testing of our fracture mechanics calculation procedure we investigate the stress in the crack region under the waterhammer action.     

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.     

压水堆主回路失水事故下水锤与管道结构的相互作用分析

传统的水锤分析和管道动力响应计算是分开的,存在一定的缺陷。本文针对核电站主回路假想双端断裂时系统的受力和力矩分析这一问题,对破裂管道分充体和管道的耦合机制,引入描述流体－管道单元的１４个参数和１４个偏微分方程,利用特征线法对水锤和管道结构的相互耦合作用进行了模拟计算。计算得到了更为准确的水锤波和管道的受力和力矩,其波形和数值均与不考虑耦合作用时有所不同。这些计算结果为压水堆核电站的核安全设计和分析     

Predicting local distributions of erosion–corrosion wear sites for the piping in the nuclear power plant using CFD models

The erosion–corrosion (E/C) wear is an essential degradation mechanism for the piping in the nuclear power plant, which results in the oxide mass loss from the inside of piping, the wall thinning, and even the pipe break. The pipe break induced by the E/C wear may cause costly plant repairs and personal injures. The measurement of pipe wall thickness is a useful tool for the power plant to prevent this incident. In this paper, CFD models are proposed to predict the local distributions of E/C wear sites, which include both the two-phase hydrodynamic model and the E/C models. The impacts of centrifugal and gravitational forces on the liquid droplet behaviors within the piping can be reasonably captured by the two-phase model. Coupled with these calculated flow characteristics, the E/C models can predicted the wear site distributions that show satisfactory agreement with the plant measurements. Therefore, the models proposed herein can assist in the pipe wall monitoring program for the nuclear power plant by way of concentrating the measuring point on the possible sites of severe E/C wear for the piping and reducing the measurement labor works.     

Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina Nuclear Power Plant

Stress corrosion cracks have been discovered in Group Distribution Headers (GDH) at the Ignalina and Chernobyl Nuclear Power Plants. This increases the probability that a guillotine pipe break can occur that creates a whipping pipe (GDH) with the potential to damage surrounding structures—i.e. adjacent GDH and its attached piping or adjacent reinforced concrete compartment wall. The GDH is the most important component for reactor safety in case of an accident. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the ECSS into the GDH.Presented in this paper is the transient analysis of a Group Distribution Header following a guillotine break at the blind end of the header. Using a very conservative force loading function, the transient response of a whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is obtained using finite element methodology.The results of the study, assuming that the impacted GDH does not suffer stress corrosion cracking, indicate that the structural integrity of the compartment should be maintained and failure should not propagate from GDH to GDH.     

Research for the rationalization of nuclear power plant pipe break criteria

The need for a new design basis for pipe break criteria is demonstrated by noting the potential deleterious effect of present criteria in piping during normal operation. Recent advances in fracture mechanics and stress analysis permit development of rational, realistic and conservative criteria that will make possible significant improvements in piping system design. Research needed to form the basis for new criteria is suggested and the nuclear industry is encouraged to work towards this goal.     

管道系统的功能性评定准则研究

管道系统的功能性是不同于管道系统压力边界完整性的一项要求，美国核管理委员会（NRC）提出了管道系统功能性的2种评定准则。为了探讨功能性评定准则的来源以及应用，通过研究经典文献中有关功能性评定准则的内容，阐述了2种评定准则的来历和依据，分析了2种功能性评定准则的特点，指出了使用功能性评定准则的注意事项。通过一个管道系统功能性评定的实例，提出2种功能性评定准则在不同的核电厂设计阶段的应用策略。对于新建的核电厂，尽量使用C级限值来保证管道系统的功能性，如果是已建造的核电厂，则可以用D级限值附加5个条件来保证管道系统的功能性。