随机徐变应力影响下重力坝时变可靠度初探

由于重力坝的随机温度场以及随机徐变应力场均为非平稳随机过程，由此所涉及的结构可靠度是非平稳向量随机过程非线性组合的时变可靠度．本文计入重力坝徐变应力场和混凝土强度随机性的影响，将时变可靠度的数值解法与随机有限元法相结合，模拟施工过程，对一典型重力坝的时变可靠度进行了计算，得出了一些有益的结论．     

某微型起爆器引起的一起误爆事故分析

该文分析某微型起爆器引起的一起误爆事故，指出该起爆器设计上的缺陷及其对作业安全的威胁，提出防患建议。     

珠江三角洲土坝和堤围震害初步预测

本项预测采用概率方法,即首先根据广东省水利部门提出的55座土坝和59个重点的座标(见附图),进行概率地震危险性分析,给出它们在一定预测年限T(取50年和100年)的地震烈度Ⅰ的超越概率;其次根据近年我国发生的几次强震时处于不同烈度区内的土坝和堤防的震害资料,形成表示土坝和土堤地震易损性的震害概率矩阵;然后,根据上述两方面结果用概率公式评价各土坝和堤围点在预测年限内各震害等级的超越概率,并取10％超越概率为标准,估计震害等级。结果表明,珠江三角洲土工构筑物的地震危害性南部高于北部。     

大坝施工混凝土质量的评价体系

本文介绍了大坝施工混凝土质量的评价系统软件的设计思路和设计成果。它利用现代计算机技术和应用模糊数学理论评价混凝土重力坝和拱坝单块和整体的质量。在采用模糊综合评价工程质量中，对不同工艺环节，对坝体不同部位，鉴于其质量要求的差异应赋予不同的权重，在权重的取值上采用了专家调查排序，定量转化的方法是比较适宜的。在程序设计中，评价数学模型采用了BASIC程序，访膜块通过*TXT文件实现了与整个大坝混凝土质量微机管理系统的接口及数据传递。操作运行方便灵活。     

煤矿许用毫秒延期电雷管研究

文中对影响煤矿许用毫秒延期电雷管沼气安全性与延期时间精度的因素进行了研究。应用正交设计研制了新的延期药配方，确定了合理的雷管结构及装配工艺参数，产品性能达到ＧＢ８０３１《工业电雷管》的规定。     

Simulation of the VVER-1000 pump start-up experiment of the OECD V1000CT benchmark with CATHARE and TRAC-PF1

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark for VVER-1000 was defined. The benchmark consists of calculation of a pump start-up experiment labelled V1000CT-1 (Phase 1), as well as a vessel mixing experiment and main steam line break (MSLB) transient labelled V1000CT-2 (Phase 2), respectively. The reference nuclear plant is Kozloduy-6 in Bulgaria. The overall objective is to assess computer codes used in the analysis of VVER-1000 reactivity transients. A specific objective is to assess the vessel mixing models used in system codes. Plant data are available for code validation consisting of one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of CATHARE and TRAC-PF1 system code results for V1000CT-1, Exercise 1, which is a full plant point kinetics simulation of a reactor coolant system (RCS) pump start-up experiment. The reference plant data include integral and sector average parameters. The comparison is made from the point of view of vessel mixing and full system simulation. CATHARE used a six-sector multiple 1D vessel thermal-hydraulic model with cross flows and TRAC used a six-sector, 18-channel coarse-mesh 3D vessel model. Good agreement in terms of integral parameters and inter-loop mixing is observed.     

动态模糊神经网络在大坝变形预报中的应用

针对静态模糊神经网络的局限性,提出了在线动态建模的模糊神经网络方法.当新增样本进入训练集之后,根据新样本对模型的贡献大小,在已有模型的基础上进行动态修正,这样可以减少建模的计算时间.新方法实现了增加样本而矩阵阶数不增加,避免了矩阵求逆运算,理论上可以提高计算效率.实例表明动态模糊神经网络方法是可行的,可实现持久预报,具有较强的适应能力和较高的预报精度,可应用于在线实时变形预报及相关领域.     

浅谈小型水库安全管理现状及对策

济南市长清区小型水库运行管理中发现了一些值得探讨的问题.本文主要介绍了济南市长清区小型水库安全运行管理存在的主要问题,分析了问题产生的原因,提出了相应的处理对策,以供参考.     

Application of condition-based HRA method for a manual actuation of the safety features in a nuclear power Plant

A practical approach to develop a more realistic fault-tree model with a consideration of various conditions endured by a human operator is proposed. In safety-critical systems, the generation failure of an actuation signal is caused by the concurrent failures of the automated systems and an operator action. These two sources of safety signals are complicatedly correlated. The failures of sensors or automated systems will cause a lack of necessary information for a human operator and result in error-forcing contexts such as the loss of corresponding alarms and indications. It is well known that the error-forcing contexts largely affect the operator's performance. An automated system which consists of multiple processing channels and complex components is also affected by the availability of the sensors. This paper proposes a condition-based human reliability assessment (CBHRA) method in order to address these complicated conditions in a practical way. We apply the CBHRA method to the manual actuation of the safety features such as a reactor trip and auxiliary feedwater actuation in Korean Standard Nuclear Power Plants. Even the human error probability of each given condition is simply assumed, the application results prove that the CBHRA effectively accommodates the complicated error-forcing contexts into the fault trees.