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。     

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管道钢的随机循环应变-寿命关系

试验研究了管道钢0Cr18Ni10Ti的随机循环应变-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-应变-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。     

变温法测量贮氢材料-氢系P-C-T曲线

描述了采用变温法测量室温下平衡压为10^-3～10^-5Pa量级、易于毒化的贮氢材料－氢系P－C－T曲线的测量原理、装置和具体测试步骤等，并通过测量贮氢材料－氢系P－C－T曲线检验了变温法的操作可行性，同时对变温法的特点和存在的问题进行了初步分析。     

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。     

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。     

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。     

ZrV2合金的吸、放氘性能研究

测定了ZrV2合金的吸、放氘P－C－T曲线、吸附（或离解）平衡压与温度的关系曲线、初始吸附速率曲线，确定了ZrV2合金的吸、放氘热力学参数，给出了ZrV2合金的吸附容量、室温吸附（或离解）平衡压，为该材料的应用提供了依据。     

用三种模型处理二元离子交换动力学的对比研究

章以强酸性阳离子交换树脂作交换剂，用三颈瓶法研究了Ｈ＾＋－ＵＯ２＾２＋离子交换反应动力学，分别用三种动力学模型处理实验数据，计算了扩散系数并用动边界模型计算了化学反应速度常数，讨论了化学反应影响交换速度的可能性并提出了化学反应在一定条件下可以延迟扩散的观点，通过对不同实验条件下的动力学曲线和计算机模型拟结果的分析，证明用动边界模型处理该交换反应是较为合理的。     

电子储存环的红外边缘辐射

介绍了电子储存环中偏转磁铁入口和出口处的非均匀磁场产生的辐射——边缘辐射，给出了零边缘长度、有限边缘长度、有限直线节长度条件下的辐射通量和亮度分布，通过与常规同步辐射场的比较，充分说明了边缘辐射是一种优良的红外同步辐射光源。     

Simplified numerical model for defining Ledinegg flow instability margins in MTR research reactor

Establishment of safety margins and the corresponding operating condition limits will ensure achievement of a safe operation of nuclear installations. For this purpose, several critical phenomena have been analyzed theoretically and experimentally and a great number of models and correlations are made available. Among these critical issues the well-known flow instability has been intensively investigated by several authors especially for nuclear power plants' (NPPs) operating conditions. However, limited published work is available for research reactor operation conditions. In general, the Whittle and Forgan correlation is widely used to define the margin to static flow instabilities in narrow parallel heated channels for research reactors.In the framework of verification and assessment of the capabilities of the RELAP5/Mod 3 system code to determine the onset of flow instability in research reactor conditions, a simple model based on steady-state equations adjusted with drift-flux correlations has been developed. The program is used to draw the pressure drop characteristic curves and to establish the conditions of the Ledinegg instability in a uniformly heated channel subject to constant outlet pressure. The model is assessed by using experimental data from a thermal hydraulic test loop by Siman-Tov and numerical results from RELAP5/Mod 3. The model presents acceptable estimation of the target mass flow that would induce flow instability and the latter could be then used to establish a conservative margin to the Ledinegg instability.     

Numerical computation of fragility curves for NPP equipment

The seismic probabilistic risk assessment (PRA) methodology is a popular approach for evaluating the risk of failure of engineering structures due to earthquake. In this framework, fragility curves express the conditional probability of failure of a structure or component for a given seismic input motion parameter A, such as peak ground acceleration (PGA) or spectral acceleration. The failure probability due to a seismic event is obtained by convolution of fragility curves with seismic hazard curves. In general, a log-normal model is used in order to estimate fragilities. In nuclear engineering practice, these fragilities are determined using safety factors with respect to design earthquake. This approach allows to determine fragility curves based on design study but largely draws on expert judgement and simplifying assumptions. When a more realistic assessment of seismic fragility is needed, simulation-based statistical estimation of fragility curves is more appropriate. In this paper, we will discuss statistical estimation of parameters of fragility curves and present results obtained for a reactor coolant system of nuclear power plant. We have performed non-linear dynamic response analyses using artificially generated strong motion time histories. Uncertainties due to seismic loads as well as model uncertainties are taken into account and propagated using Monte Carlo simulation.     

Development of multicomponent vaporization/condensation model for a reactor safety analysis code SIMMER-III: Theoretical modeling and basic verification

It is believed that the numerical simulation of thermal-hydraulic phenomena of multiphase, multicomponent flows in a reactor core is essential to investigate core disruptive accidents (CDAs) of liquid-metal fast reactors. A new multicomponent vaporization/condensation (V/C) model was developed to provide a generalized model for a fast reactor safety analysis code SIMMER-III, which analyzes relatively short-time-scale phenomena relevant to accident sequences of CDAs. The model characterizes the V/C process associated with phase transition through heat-transfer and mass-diffusion limited models to follow the time evolution of the reactor core under CDA conditions. The heat-transfer limited model describes the nonequilibrium phase-transition processes occurring at interfaces, while the mass-diffusion limited model is employed to represent effects of noncondensable gases and multicomponent mixture on V/C processes. Verification of the model and method employed in the multicomponent V/C model of SIMMER-III was performed successfully by analyzing a series of multicomponent phase-transition experiments.     

反应堆压力容器辐照脆化状态评估

基于大量相似辐照脆化试验测试数据和实际辐照监督测试数据,采用统计分析的方法,选出适合于某核电厂反应堆压力容器（RPV）的辐照脆化评估公式。以该核电厂已经完成的辐照监督管测试数据为输入,对RPV当前的辐照脆化状态进行了评估,并推算、分析了RPV在寿期末的结构完整性;基于辐照脆化计算结果,绘制了各运行阶段RPV的压力-温度限值曲线（P-T曲线）,并给出运行建议。      

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.     

0Cr18Ni10Ti管道钢的概率机械性能

完成了反应堆管道新材料0Crl8Nil0Ti的概率机械性能试验研究。拓展赵等提出的确定有限疲劳可靠性数据良好假设分布的统一方法到本文机械性能数据分析。结果表明．数据具有明显正偏性，常见正态分布描述不适于该套数据。综合考虑各分布的拟合优度、与失效机制的一致性和尾部预测的安全性．数据的良好统计分布是对数正态分布和极大值分布。从统计分析的方便性角度。提出了给定可靠度和置信度下基于对数正态分布的机械性能参数估计方法。并有效地估计了材料的概率机械性能参数。     

A statistical approach to the prediction of pressure tube fracture toughness

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in a nuclear reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The paper presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower-bound approach.     

Development of a thermal-hydraulic subchannel analysis code for motion conditions

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable.     

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors (IV)

A calculation model has been developed in order to evaluate effectiveness of hydrazine and hydrogen co-injection (HHC) into reactor water for mitigation of intergranular stress corrosion cracking of structural materials used in boiling water reactors (BWRs). The HHC uses the strong reducing power of hydrazine radical, which is produced in the downcomer region under irradiation by γ-rays and neutrons. Some reactions and their reaction rate constants were determined based on experiments which were carried out in aerated water, hydrogenated water, and deaerated water. The calculated results were in good agreement with experimental data by a factor of two. The model was applied to a BWR and it was found that the HHC cut oxygen and hydrogen peroxide amounts dissolved in reactor water more effectively than hydrogen water chemistry alone. Thus, the required amount of hydrogen for hydrazine injection was much lower than that for hydrogen water chemistry. Consequently, electrochemical corrosion potential of structural materials could be lowered below–0:1V vs. SHE without any increase of MS line dose rate, which has been a limitation of the conventional hydrogen water chemistry. The HHC was predicted to decrease crack growth rate of structural materials by a factor of 10.