Important factors affecting fault detection coverage in probabilistic safety assessment of digital instrumentation and control systems

As digital instrumentation and control (I&C) systems are gradually introduced into nuclear power plants (NPPs), concerns about the I&C systems’ reliability and safety are growing. Fault detection coverage is one of the most critical factors in the probabilistic safety assessment (PSA) of digital I&C systems. To correctly estimate the fault detection coverage, it is first necessary to identify important factors affecting it. From experimental results found in the literature and the authors’ experience in fault injection experiments on digital systems, four system-related factors and four fault-related factors are identified as important factors affecting the fault detection coverage. A fault injection experiment is performed to demonstrate the dependency of fault detection coverage on some of the identified important factors. The implications of the experimental results on the estimation of fault detection coverage for the PSA of digital I&C systems are also explained. The set of four system-related factors and four fault-related factors is expected to provide a framework for systematically comparing and analyzing various fault injection experiments and the resultant estimations on fault detection coverage of digital I&C systems in NPPs.     

Uncertainty analysis of CsI distribution for typical CANDU6 station blackout accident

This paper illustrates an application of a severe accident analysis code, ISAAC (Integrated Severe Accident Analysis Code for the CANDU plants), to the uncertainty analysis of fission product behaviors during a severe reactor accident. The ISAAC code is a system-level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, and whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user options for supporting sensitivity and uncertainty analyses. The present application is mainly focused on determining an estimate of the fission products in the release and transport processes and the relative importance of the dominant contributors to the predicted fission products. The key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the fission product release correlations, vapor–aerosol equilibrium, vapor–surface equilibrium for a revaporization calculation, and aerosol decontamination factors. A typical CANDU6 type plant, the Wolsong nuclear power plant, was used as a reference plant for the analysis.     

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.     

Two viewpoints for software failures and their relation in probabilistic safety assessment of digital instrumentation and control systems

As the use of digital systems in nuclear power plants increases, the reliability of the software becomes one of the important issues in probabilistic safety assessment. In this paper, two viewpoints for a software failure during the operation of a digital system or a statistical software test are identified, and the relation between them is provided. In conventional software reliability analysis, a failure is mainly viewed with respect to the system operation. A new viewpoint with respect to the system input is suggested. The failure probability density functions for the two viewpoints are defined, and the relation between the two failure probability density functions is derived. Each failure probability density function can be derived from the other failure probability density function by applying the derived relation between the two failure probability density functions. The usefulness of the derived relation is demonstrated by applying it to the failure data obtained from the software testing of a real system. The two viewpoints and their relation, as identified in this paper, are expected to help us extend our understanding of the reliability of safety-critical software.     

Development of Systematic Models for Aerosol Agglomeration and Spray Removal under Severe Accident Conditions

Radionuclide behavior during various severe accident conditions has been addressed as one of the important issues to discuss environmental safety in nuclear power plants. The present paper deals with the development of analytical models and their validations for the agglomeration of multiple-component aerosol and spray removal that controls source terms to the environment of both aerosols and gaseous radionuclides during recirculation mode operation in a containment system for a light water reactor. As for aerosol agglomeration, the single collision kernel model that can cover all types of two-body collision of aerosol was developed. In addition, the dynamic model that can treat aerosol and vapor transfer leading to the equilibrium condition under the containment spray operation was developed. The validations of the present models for multiple-component aerosol growth by agglomeration were performed by comparisons with Nuclear Safety Pilot Plant (NSPP) experiments at Oak Ridge National Laboratory (ORNL) and AB experiments at Hanford Engineering National Laboratory (HEDL). In addition, the spray removal models were applied to the analysis of containment spray experiment (CSE) at HEDL. The results calculated by the models showed good agreements with experimental results.     

乳液型丙烯酸酯压敏胶的研究

以丙烯酸丁酯(BA)、甲基丙烯酸甲酯(MMA)和丙烯酸(AA)为反应单体,过硫酸铵(APS)为引发剂,非离子型乳化剂(OP-10)/十二烷基硫酸钠(SDS)为复合乳化剂,硫酸铝为化学交联剂,采用乳液聚合法合成了丙烯酸酯PSA(压敏胶)。研究结果表明:当w(BA)=80%、w(MMA)=14%、w(AA)=6%、w(APS)=0.4%、w(硫酸铝)=2.0%、w(复合乳化剂)=6%且m(OP-10)∶m(SDS)=2∶1时,相应PSA的固含量(54.41%)较高、黏度(140 mPa.s)较低以及耐热性良好,并且其180°剥离强度(2.79 N/mm)、初粘力(16号钢球)和持粘力(>24.0 h)俱佳。     

丙烯酸酯PSA环形初粘力影响因素研究

以反应型乳化剂(DNS-86)/阴离子型乳化剂(2A1)为复合乳化剂、甲基丙烯酸(MAA)与甲基丙烯酸羟乙酯(HEMA)为极性单体和正十二硫醇为链转移剂时,采用单体预乳化法和半连续乳液聚合法制备丙烯酸酯PSA(压敏胶)乳液。考察了PSA胶带的基材、干胶厚度、烘干条件、复合乳化剂、极性单体和链转移剂等对环形初粘力的影响。结果表明:当基材为白色BOPP(双向拉伸聚丙烯)薄膜、干胶厚度为50μm、烘干时间为3 min、烘干温度为110~115℃、w(正十二硫醇)=0.09%、同时引入MAA和HEMA极性单体、w(复合乳化剂)=1.5%和m(2A1)∶m(DNS-86)=2∶1时,相应丙烯酸酯PSA乳液的环形初粘力相对最大(14.73 N/25 mm)。     

PSA/BAS水刺滤料过滤性能研究

以玄武岩纤维(BAS)机织布为底层,芳砜纶(PSA)为面层,采用水刺工艺使二者结合制得PSA/BAS水刺滤料.测试了PSA/BAS滤料的过滤性能,并与100％ PSA针刺滤料和BAS机织布进行对比,分析了PSA/BAS水刺滤料用于袋式除尘的技术特点.结果表明:在不同过滤速度下,清洁PSA/BAS水刺滤料对粒径≥6μm颗...     

UV固化型高性能压敏胶的发展动态

UV固化型压敏胶作为一种新型光固化压敏胶,无溶剂、无污染、能耗低,既克服了热熔压敏胶容易热氧老化和不耐高温的缺点,与乳液型压敏胶相比,又提高了耐水性和涂布速度,成为压敏胶领域研究的热点。本文综述了近10年来国内外UV固化型压敏胶的发展状况,对不同种类的UV固化型压敏胶的性能特点和制备方法做了归纳,重点阐述了丙烯酸酯类UV固化压敏胶的最新进展。总结了UV固化型压敏胶需要解决的问题并对未来发展方向作了展望。     

脱碳工艺对制氢成本的影响分析

本文介绍了轻烃水蒸汽转化制氢装置中几种中变气脱碳工艺,分析了制氢流程中脱碳工序的利弊。通过对10000Nm3.h-1天然气制氢装置投资和能耗计算,分析了不同脱碳工艺流程对制氢成本的影响,结果表明:根据市场情况确定脱碳工艺,可以实现经济效益的最大化。