OCr18Ni1OTi管道钢的随机循环应变-寿命关系

试验研究了管道钢OCr18Ni1OTi的随杨循环应变-寿命关系。基于Coffin-Manson方程，提出了考虑了任意存活概率和置信度的随机CSL关系的模型及参数的求解方法模型由概率-应变-寿命曲线、置信度-寿命曲线和概率-置信度-应变-寿命曲线组成，分别用于表征试验数据分散性规律、数据量以及两者同时对概率评价的影响。试验数据的分析结果验证了模型的有效性和实用性。     

0Cr18Ni10Ti管道钢的随机循环本构模型

通过完成增量步应变控制疲劳试验,研究了新管道钢0Cr18Ni10Ti的随机循环本构关系。试验验证了以前在焊缝金属试验中的发现与推断,即工程材料的循环本构存在随机性,与循环应变-寿命关系的随机性一样,是固有的疲劳现象。拓展以前赵等人的工作(Nucl. Eng. Des., 2000, 199(3): 315-326),基于Ramberg-Osgood方程及其修正形式,提出了任意存活概率和置信度的随机循环本构模型及参数的求解方法。模型包括存活概率-应变-寿命曲线、置信度-应变-寿命曲线和存活概率-置信度-应变-寿命曲线。试验数据分析验证了模型的有效性和实用性。     

0Crl8Ni10Ti管道钢的随机循环本构模型

通过完成增量步应变控制疲劳试验，研究了新管道钢0Cr18Ni10Ti的随机循环本构关系．试验验证了以前在焊缝金属试验中的发现与推断，即工程材料的循环本构存在随机性．与循环应变-寿命关系的随机性一样，是固有的疲劳现象。拓展以前赵等人的工作(Nucl．Eng．Des．，2000，199(3)：315-326)．基于Ramberg-Osgood方程及其修正形式．提出了任意存活概率和置信度的随机循环本构模型及参数的求解方法．模型包括存活概率．应变．寿命曲线、置信度．应变．寿命曲线和存活概率-置信度-应变-寿命曲线试验数据分析验证了模型的有效性和实用性。     

16MnR钢焊接头概率循环应变-寿命模型的综合分析

对16MnR钢焊接头概率循环应变一寿命模型作了综合分析。综合利用不同试验条件下的试验数据,提出了合理的材料与结构概率设计分析曲线。模型用概率三参数幂指数模型表征,同时考虑了试验数据分散性和样本量对概率评价的影响,包含了存活概率曲线、置信度曲线和两者融合曲线,可为工程实践提供广泛的选择性。作为比较,相应给出了16MnR钢母材的各曲线参量。     

有限数据下结构疲劳可靠性设计曲线的统计模型

提出了有限数据下确定反应堆结构疲劳可靠性设计S－N曲线的统计模型，该模型置信度－存活率概率－强度－寿命（C－P－S－N）曲线来表示考虑了常规P－S－N曲线预测结果的可信性，应用线性回归统计分析原理，将置信度引入到常规P－S－N曲线中，从而得到C－P－S－N曲线。1Cr18Ni9Ti不锈钢管道焊接缝金属材料试验数据分析结果说明了模型的合理性和可用性，同时还说明了有限数据下常规P－S－N曲线偏于危险。     

结构材料常规疲劳数据库开发

开发了反应堆结构材料常规疲劳数据库管理系统，库中包含了常规疲劳设计分析曲线参量，即应力－寿命、概率－应力－寿命、应变－寿命、概率－应变－寿命、循环应力－应变及概率－循环应力－应变曲线参量，以及相关材料和试验条件信息。可通过材料名称或机械性能区间指标查询。根据查询结果，可进行常规疲劳设计分析与评价或常规疲劳可靠性设计分析与评价。     

平均应变对0Cr18Ni10Ti管道钢随机应变-疲劳寿命关系的影响规律

试验研究了平均应变对0Cr18Ni10Ti管道钢随机应变-寿命关系的影响规律从节约试样和试验费用角度,采用改进的极大似然疲劳试验法,应变比分别为-1、-0．52、-0．22、0．029、0．18和0．48的条件下,完成了104个试样的应变控制疲劳试验在材料具有完全平均应力松弛特征、现有平均应变理论无法表征其影响情况下,首先基于Coffin—Manson方程,应用广义极大似然法有效地测定出各应变比下材料的随机应变-寿命关系．通过比较各应变比下疲劳寿命均值、对数疲劳寿命均方差和概率疲劳寿命,揭示出0Cr18Ni10Ti管道钢高可靠性时,应变比大于0有大约1．3-1．6的正面效应,小于0为负面效应现有平均应变效应研究仅考虑疲劳寿命均值可能给出错误评价;必须综合考虑均值、均方差和样本量3因素才能给出合理评价。     

估计三种常用应力 -寿命模型概率设计 S-N曲线的统一方法

提出了适于三参数、Langer和Basquin三种常用应力－寿命模型称为广义极大似然法的估计概率设计S－N曲线及其置信限的统一方法。方法将概率设计S－N曲线表示为对数疲劳寿命均值和均方差曲线的广义形式。与现有常规和经典极大似然法方法不同,考虑所有试验数据的统计特征,应用最小二乘法先估计出均值曲线中的材料常数,然后利用极大似然原理和数学规划法估计出均方差曲线中的材料常数。有效性采用拟合相关系数、拟合误差均方差值和置信限综合评价。对反应堆不锈钢管道焊接头虚拟应力幅－裂纹萌生寿命数据及45＃碳钢成组法和极大似然法疲劳试验应力－寿命数据的分析说明了方法的有效性。一般来说,三参数模型的拟合效果最好,Langer模型次之,Basquin模型最差。本文方法的拟合效果好于现有方法,并尽量避免了现有方法受试验数据局部统计特征影响而可能给出偏于非安全估计的缺陷。     

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。     

平均应变对0Gr18Ni10Ti钢随机循环应力-应变关系的影响规律

试验研究了平均应变对0Cr18Ni10Ti管道钢随机循环应力-应变关系的影响规律.从节约试样和试验费用角度,提出了改进的极大似然疲劳试验法,以应变比分别为-1、-0.52、-0.22、0.029、0.18和0.48完成了104个试样的应变控制疲劳试验.揭示出材料具有Masing行为和完全平均应力松弛特征,现有平均应变理论无法表征其影响.因此,首先基于Ramberg-Osgood方程,应用广义极大似然法有效地测定出各应变比的材料随机循环应力-应变关系.通过比较各应变比的循环应力幅均值、循环应力幅均方差和概率循环应力幅,揭示出高可靠性管理时,平均应变具有降低循环应力幅作用,应变比接近0时最显著,偏离逐渐减弱.说明了研究平均应变效应,仅考虑均值,可能给出错误评价;必须综合考虑均值、均方差和样本量3因素才能给出合理评价.     

How to eliminate non-damaging earthquakes from the results of a probabilistic seismic hazard analysis (PSHA)—A comprehensive procedure with site-specific application

The results of probabilistic seismic hazard analyses are frequently presented in terms of uniform hazard spectra or hazard curves with spectral accelerations as the output parameter. The calculation process is based on the evaluation of the probability of exceedance of specified acceleration levels without consideration of the damaging effects of the causative earthquakes. The same applies to the empirical attenuation equations for spectral accelerations used in PSHA models. This makes interpreting and using the results in engineering or risk applications difficult. Uniform hazard spectra and the associated hazard curves may contain a significant amount of contributions of weak, low-energy earthquakes not able to damage the seismically designed structures of nuclear power plants. For the development of realistic engineering designs and for realistic seismic probabilistic risk assessments (seismic PRA) it is necessary to remove the contribution of non-damaging earthquakes from the results of a PSHA. A detailed procedure for the elimination of non-damaging earthquakes based on the CAV (Cumulative Absolute Velocity)-filtering approach was developed and applied to the results of the large-scale PEGASOS probabilistic seismic hazard study for the site of the Goesgen nuclear power plant. The procedure considers the full scope of epistemic uncertainty and aleatory variability present in the PEGASOS study. It involves the development of a set of empirical correlations for CAV and the subsequent development of a composite distribution for the probability of exceedance of the damaging threshold of 0.16 gs. Additionally, a method was developed to measure the difference in the damaging effects of earthquakes of different strengths by the ratio of a power function of ARIAS-intensity or, in the ideal case, by the ratio of the square roots of the associated strong motion durations. The procedure was applied for the update of the Goesgen seismic PRA and for the confirmation of a revised safe shutdown earthquake for the Goesgen nuclear power plant. The application of the procedure leads to results which are in reasonable compliance with evaluations based on the macroseismic method using European macroseismic intensities and associated vulnerabilities. The paper is an extended version of the paper #1142 presented at the 19th SMIRT conference in Toronto, 2007.     

设计S—N曲线的概率估计和ASME法的可靠性评价

基于ＬａｎｇｅｒＳ－Ｎ曲线,提出了设计Ｓ－Ｎ曲线的概率估计方法和ＡＳＭＥ法的可靠性评价方法。Ｐ－Ｓ－Ｎ曲线由文中提出的广义极大似然法估计。方法可应用于处理具有双随机性特征的Ｓ－Ｎ数据,并可推广应用于处理应力控制成组法疲劳试验和极大似然法试验得到的Ｓ－Ｎ数据。     

Prioritization of components important for safety and sensitive to ageing. Application for TRIGA reactor

Due to potential impact on the SSC performances, the identification of ageing effects and implementation of appropriate methods for mitigation of these effects represents an important preoccupation of many organizations and has received even more attention in the last years in the perspective of Long Term Operation. The efforts used for developing time-dependent reliability models for all SSC could be considerable and may outweigh the benefits, but they are not always necessary, mainly because in some cases, the maintenance, test, inspection and surveillance methods are good provision to mitigate the ageing effects. To efficiently use the limited resources, it is necessary to identify and prioritize the SSC that need time-dependent reliability models by using specific criteria. This paper contributes to the effort of performing an efficient process of evaluation of ageing effects using probabilistic safety assessment models. The paper presents the approach developed for SSC selection, and the results of its application carried out for two particular systems of the TRIGA research reactor.     

Treatment of uncertainties in reliability models

For systems in Nuclear Installations it is necessary to analyze the possible mechanisms for failure of each component, to perform probabilistic analysis for the expected rate of such failure and to aggregate the component failure logically and numerically to a prediction of the reliability of the whole system. In this process we must treat models and data associated with uncertainties of different kinds. Also subjective judgements are necessary for an estimation of these uncertainties in the models and data. Based on a system example this paper illustrates some of these aspects more detailed.     

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.     

Some thoughts on the future of probabilistic structural design of nuclear components

This paper presents some views on the future role of probabilistic methods in the structural design of nuclear components. The existing deterministic design approach is discussed and compared to the probabilistic approach. Some of the objections to both deterministic and probabilistic design are listed. Extensive research and development activities are required to mature the probabilistic approach sufficiently to make it cost-effective and competitive with current deterministic design practices. The required research activities deal with probabilistic methods development, more realistic causal failure mode models development, and statistical data models development. A quasi-probabilistic structural design approach is recommended which accounts for the random error in the design models.