对法国高放废物处置安全审评的分析研究

介绍了法国高放废物处置研究现状和规划,对法国高放处置场的审评技术单位法国核与辐射安全研究院（IRSN）所开展的高放处置安全研究和审评工作及其提出的审评原则和审评要点进行了分析研究,并对我国的高放处置安全审评工作提出了建议。     

ASME规范附录Ⅷ在中国应用的若干问题及建议

根据参加的相关审评和监督活动的经验，说明在审评和监督过程中所采取的思路和策略，同时也澄清了就检验方法替代申请所产生的若干问题，并提出了一些建议。     

换料报告审评方法及效果

换料报告是营运单位在换料停堆前向国家核安全局提交的综合报告,主要描述换料停堆期间将要进行的所有维修,试验,在役检查,装卸料等活动。通过对该报告的审评来验证营运单位所计划的活动是否满足核安全规范的要求及核电厂安全运行的需要。本文介绍了国家核安全局对大亚湾核电站两台机组第三次停堆换料报告审评时所采用的方法及审评中发现的问题。     

岭澳核电厂先进燃料管理项目的审评

本文简要描述了岭澳核电厂先进燃料管理项目的审评工作，介绍了审评范围和审评中重点问题的审评过程，并给出了审评结论。     

在役核辅助管道焊缝缺陷事件的核安全审评

文章给出了一个核安全审评的实例,即某核电厂1号机组核辅助管道焊缝缺陷事件的审评;分析了缺陷产生的原因、可能存在缺陷焊缝的范围以及处理措施;探讨了焊缝缺陷事件所暴露出的我国核安全设备设计、制造、安装活动中存在的问题.     

核动力厂温排水环境影响审评原则的讨论

简要描述了制定核动力厂温排水环境影响审评原则的必要性,介绍了审评原则的编制过程中所参考的国外实践和技术基础以及审评原则的主要内容,并对审评原则中的热点问题进行了讨论。     

核电厂设备抗震鉴定的审评

介绍了核电厂设备抗震鉴定的标准和要求，鉴定的范围、步骤、方法和程序，以及抗震鉴定审评所依据的准则。详细论述了抗震试验鉴定中所采用的具体实施程序，并通过审评实践对核电厂设备抗震鉴定中常见的一些问题也进行了探讨。     

地震动力分析中的输入问题

陈述了地震动力分析的核安全审评依据，分析了在民用核设施的安全审评中关于地震动力分析所遇到的各种不同情况，其中包括支撑介质类型（如岩石地基、非岩石地基、深厚软土层地基等）以及所遵循的规范标准（如中国规范、法国规范、美国规范等）。给出了其中各种典型实例的自然状况以及就地震动力分析的输入问题的审评经验和体会。笔者认为，在地震输入问题上和在ＨＡＦ０１０１（１）的执行过程中，尚存在着值得探讨的地方。     

EPR核电厂设备安全分级审评和评述

介绍了EPR核电厂设备安全分级方法,阐述了台山核电厂审评中的主要观点,对比M310机组,评述了EPR核电厂设备安全分级问题审评工作,对今后EPR机组的安全审评提出了建议。     

核电厂、研究堆及专用核设施事故分析审评模式探讨

对核电厂、研究堆和专用核设施安全分析报告的核安全审评。是为了克保核电厂、研究堆和专用核设施在正常运行和各类事故条件下的安全,保护工作人员、公众和环境。其中,对安全分析报告中事故分析的审评,是整个核安全审评的关键,本文结合作者近年来对部分核电厂,研究堆和专用核设施的审评实践,对事故分析审评模式进行初步的探讨,作者认为对事故分析的审评,应该遵循有关的核安全法规和规范,着重关注所分析的事故谱是否完整、是否满足事故分析的基本假定、初台条件是否保守、计算机程序是否适用、包络性事故分析的结果是否满足验收准则等方面,此外,事故分析应借鉴国际核安全实践经验与技术发展,数据与其它章节要自洽,分析要考虑人因工程,分析结果应合理可信。我国目前还没有一套关于核设施事故分析的规范和标准,作者对制定这方面的法规和规定提出了自己的看法,通过对事故分析审评模式的探讨,能促进审评者和营运者更好理解法规,也能使双方在法规和现实之间及时达成一致意见,对我国的核事业健康发展是有积极作用的。     

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.

For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems.     

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR⁺ (ASBWR⁺) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Issues affecting advanced passive light-water reactor safety analysis

Next generation commercial reactor designs emphasize enhanced safety by means of improved safety system reliability and performance. These objectives are achieved via safety system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet, the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs will necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing U.S. advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes may require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.     

先进压水堆非能动安全系统研究进展

介绍了我国先进压水堆非能动安全系统研究进展及国内外先进压水堆非能动安全系统研究发展状况,指出我国非能动安全系统研究的发展方向是进行新一代1000MW级压水堆非能动安全系统的研究。     

Safety approach of DFBR design study in Japan

The goal of the safety design for the demonstration fast breeder reactor is to ensure that the safety level is equivalent to or higher than that of the light water reactors of the same period. The design of the safety features such as reactor shutdown, decay heat removal and confinement systems is of importance to reach the goal. The reactor core is equipped with two independent fast shutdown systems, the primary system and the backup system. In addition, it is planned to strengthen the passive shutdown capability by using self- actuated systems such as a Curie point device for the backup system. The decay heat is removed from the core to the atmosphere through the safety lines of the direct reactor auxiliary cooling system which is composed of four independent lines. Furthermore, under the severe conditions that no active function of the decay heat removal system is available, the heat can be removed by natural convection through the safety lines by taking advantage of the high boiling temperature of sodium. For the confinement function, the reactor vessel is surrounded by a containment vessel and a confinement area.

The design concept of these safety features is described in this paper.     

Safety evaluation of the inherent and passive safety features of the smart design

SMART (system-integrated modular advanced reactor) is a 330 MWt advanced integral PWR, which is under development at KAERI for seawater desalination and electricity generation. The conceptual design of the SMART desalination plant produces 40,000 m³/day of potable water and generates about 90 MW of electricity, which are assessed as sufficient for a population of about 100,000. The SMART enhances safety by adopting the inherent safety design features such as the elimination of large break loss of coolant accidents, substantially large negative moderator temperature coefficients, etc. In addition, the safety goals of the SMART are achieved through the adoption of passive engineered safety systems such as an emergency core cooling system, passive residual heat removal system, safeguard vessel, and reactor and containment overpressure protection systems. This paper describes the design concept of the major safety systems of the SMART and presents the results of the safety analyses using a MARS/SMR code for the major limiting accidents including transient behaviors due to desalination system disturbances. The analysis results employing conservative initial/boundary conditions and assumptions show that the safety systems of the SMART conceptual design adequately remove the core decay heat and mitigate the consequences of the limiting accidents, and thus secure the plant to a safe condition.     

Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1

A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions.     

Reliability assessment method for NPP digital I&C systems considering the effect of automatic periodic tests

Since digital technologies have been improved, the analog systems in nuclear power plants (NPPs) have been replaced with digital systems. Recently, new NPPs have adapted various kinds of digital instrumentation and control (I&C) systems. Even though digital I&C systems have various fault-tolerant techniques for enhancing the system availability and safety compared to conventional analog I&C systems, the effects of these fault-tolerant techniques on system safety have not been properly considered yet in most probabilistic safety assessment models. Therefore, it is necessary to develop the safety evaluation method for digital I&C systems with consideration of fault-tolerant techniques. Among the various issues in the safety model for digital I&C systems, one of the important issues is how to exclude the duplicated effect of fault-tolerant techniques implemented at each hierarchy level of the system. The exact relation between faults and fault-tolerant techniques should be identified in order to exclude this duplicated effect. In this work, the relation between faults and fault-tolerant techniques are identified using fault injection experiments. As an application, the proposed method was applied to a module of a digital reactor protection system.     

Nuclear Power Plant Compliance with the Principle of Power-System Indepedence from Safety Systems Operation

The purpose of this paper is to show that during the operation of safety systems at nuclear power plants the principle of independence from the power system, which is one of the basic principles inocrporated in the design of safety systems, is not satisfied and the power system, especially if it is deficient, cannot guarantee the required electricity and protection for safety systems from general failures. To satisfy the independence principle, guarantee the required quality of electricity, and protect the safety systems in nuclear power plants from general failures, it is proposed that the presently operative algorithm for starting up diesel generators be reexamined. When the safety systems at nuclear power plants perform their required functions, they should operate from autonomous diesel generators at the nuclear power plant, which are equipped with electricity quality regulators (frequency and voltage), and not from the power system. It is also suggested that the variant of the algorithm where diesel generators are started up as a preventative measure when the quality of the electricity in the power system drops below admissable limits be reexamined.     

基于腐蚀性能的碳钢管道使役时间评价方法

核电厂管道系统腐蚀程度监测和预测使役时间是核电厂安全性能评估的重要内容之一.文章介绍了采用M（o）ssbauer谱分析碳钢管道腐蚀产物,超声无损检测确定管壁剩余厚度,依据统计学原理,计算各类管道系统管壁腐蚀程度的分布和理论腐蚀速率,用于评估管道在下一生产周期的安全状态和使役寿命.