101重水研究堆安全关闭过渡期放射性源项调查

针对我国第1座研究性反应堆（101重水研究堆）安全关闭过渡期的放射性源项调查,采用对可达部位取样分析与理论计算相结合的方法,给出了堆本体主要部件的中子活化源项。采用现场测量和对管道、设备内壁取样的方法获取了回路系统污染源项。另外,对反应堆厂房构筑物地面和墙面的污染水平、乏燃料保存水池和废树脂等进行了较为全面的现场测量和取样分析。通过源项调查,初步掌握了101重水研究堆退役的主要放射性源项的特点和存留量。     

多堆核电场址概率风险评价中的始发事件分析

绝大多数核电厂概率风险评价均以单个反应堆为单位,安全目标的制定也是如此。事实上,同一场址（site）上有多座核电厂（plant）、同一厂址（plant）上有多个反应堆（reactor）的情况并不罕见。为将概率风险评价从单个反应堆推广至整个场址,本文在定义场址风险的基础上,针对始发事件展开分析,给出其分类及识别方法建议。分析表明,多堆场址的始发事件可归入2类单堆始发事件,以及3类多堆始发事件。此结果是开展多堆场址概率风险评价的第1步,具有重要价值。     

多方法融合的反应堆紧急停堆子系统安全性分析

针对反应堆紧急停堆子系统,将故障模式影响分析（FMEA）、故障树分析（FTA）、系统理论的过程分析（STPA）3种独立的基本分析方法进行组合,形成仪表控制系统设计阶段的失效和故障基本项覆盖统计表格。STPA方法能够很好地弥补了FMEA和FTA方法的不足。同时,在仪控系统的设计阶段,STPA方法非常适合发现反应堆紧急停堆子系统涉及的软件类、系统交互以及通信类的故障和安全问题。      

研究堆的退役

【英国《国际核工程》2003年12月刊报道】 世界上建造和投入运行的研究反应堆已达数百座。目前,随着已关闭的反应堆的不断增加,其数量已与运行中的反应堆不相上下,它们的退役工作也随之成为一个重大的国际问题。由于研究堆与其他核设施相比的独特之处,因此有必要特别关注研究堆的退役。研究堆退役活动不同于其他核设施的主要方面有:研究堆(包括原型反应堆)种类繁多、研究内容也千差万别、以及一些研究堆临近公众社区。特别是,大部分国家在设计、建造、运行和关闭反应堆的时候,没有充分关注和规划好以后的退役工作。大部分国家的退役计划顶多…     

俄完成一座研究堆退役 并启动放化研究中心的建设

正【本刊2014年10月综合报道】位于季米特洛夫格勒的俄罗斯反应堆研究所(NIIAR)近期完成了两项重要工作,即完成了研究堆RBT-10/1的退役工作和启动了多用途放化研究中心的建设。研究堆退役反应堆研究所于2014年9月1日表示,俄罗斯联邦生态、技术与核能监督局(Rostechnadzor)已于8月将RBT-10/1从受     

英国明确两种可靠的处置方案

【世界核新闻网站2014年1月21日报道】英国核退役管理局（NDA）近日得出结论,Prism快堆和改进型坎杜6反应堆（EC6）都是可用于管理英国库存钚的“可靠方案”。但是,政府的优选方案仍然是以混合氧化物（MOX）燃料形式对钚进行再利用。Prism是一种311MWe的钠冷快堆。EC6是一种700MWe的重水慢化与冷却压力管式反应堆,是成熟的坎杜6（Candu6）设计的升级版。     

秦山一期堆本体退役源项估算及辐射场可视化

针对我国秦山一期核反应堆实际情况,利用蒙特卡罗程序建立了细化到燃料棒结构的全堆芯pinby-pin模型进行中子输运计算,并对计算模型的可靠性进行了验证;基于堆本体结构部件的几何参数、材料参数及堆本体中子注量率分布,在假定功率运行史的情况下,利用燃耗计算程序计算了反应堆停堆后的中子活化产物作为堆本体退役源项的估算结果,并对源项产生的三维辐射场剂量分布情况进行了可视化建模与分析,模拟结果与理论分析一致。本研究是下一步建立我国秦山核电厂退役技术安全验证和虚拟仿真平台的关键性基础工作。     

基于R2S方法的反应堆结构材料活化剂量计算研究

反应堆结构材料在堆芯中子辐照下由于中子活化反应而产生大量的放射性核素,其衰变光子是反应堆停堆检修、换料、退役过程中工作人员职业照射剂量的重要来源。本文基于严格两步法（R2S）,研究了反应堆结构材料栅元活化计算方法,并基于蒙卡粒子输运程序（MCNP）与点活化计算程序（ORIGEN）建立了反应堆结构材料活化剂量计算软件（MOCA）。通过开发功能接口与数据接口程序实现输运程序与活化计算程序的自动耦合,进而实现“中子输运-活化分析-剂量计算”全自动耦合分析。利用M5包壳活化计算模型、不锈钢活化计算模型和NUREG/CR-6115压水堆模型对MOCA进行基准验证,证明了MOCA的正确性与可靠性。      

SPRR-300堆生物屏蔽层活化情况调查

反应堆生物屏蔽层是反应堆退役阶段的重要源项之一。通过建立计算模型,使用MCNP和ORIGEN2程序计算获得了SPRR-300堆生物屏蔽层的活化情况。为校核理论计算结果,对SPRR-300堆生物屏蔽层混凝土进行了取样分析。取样分析结果与理论计算结果较为一致,证明了理论计算模型的准确性。通过对取样分析结果进行拟合,获得了SPRR-300堆生物屏蔽层针对60 Co核素的活化厚度,即1.1m高度处的活化厚度为840mm,1.5m高度处的活化厚度为680mm。     

HWRR屏蔽层活化源项研究

反应堆屏蔽层通常由钢筋混凝土浇筑而成,体积及重量巨大,是反应堆退役源项的重要来源之一。通过建立反应堆3D计算模型,利用MCNP和ORIGEN活化计算程序计算了重水研究堆(HWRR)屏蔽层不同位置的中子注量率和活化源项。为验证计算模型和计算结果的准确性,在HWRR屏蔽层活性区中央位置沿水平方向进行钻孔取样,对获得的混凝土样品中的~(60)Co和~(152)Eu的活度进行了测量,分析结果与计算结果较吻合,证明了理论计算模型的准确性。最后对HWRR屏蔽层的活化深度进行了计算,得出反应堆屏蔽层活化深度最大值为600 mm。计算结果证明保留外层屏蔽层的退役方案从理论上是可行的。     

Risk assessment on hazards for decommissioning safety of a nuclear facility

A decommissioning plan should be followed by a qualitative and quantitative safety assessment of it. The safety assessment of a decommissioning plan is applied to identify the potential (radiological and non-radiological) hazards and risks. Radiological and non-radiological hazards arise during decommissioning activities. The non-radiological or industrial hazards to which workers are subjected during a decommissioning and dismantling process may be greater than those experienced during an operational lifetime of a facility. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities and as well as during accidents. The risk assessment method was developed by using risk matrix and fuzzy inference logic, on the basis of the radiological and non-radiological hazards for a decommissioning safety of a nuclear facility. Fuzzy inference of radiological and non-radiological hazards performs a mapping from radiological and non-radiological hazards to risk matrix. Defuzzification of radiological and non-radiological hazards is the conversion of risk matrix and priorities to the maximum criterion method and the mean criterion method. In the end, a composite risk assessment methodology, to rank the risk level on radiological and non-radiological hazards of the decommissioning tasks and to prioritize on the risk level of the decommissioning tasks, by simultaneously combining radiological and non-radiological hazards, was developed.     

Factor analysis on hazards for safety assessment in decommissioning workplace of nuclear facilities using a semantic differential method

The decommissioning of nuclear facilities must be accomplished according to its structural conditions and radiological characteristics. An effective risk analysis requires basic knowledge about possible risks, characteristics of potential hazards, and comprehensive understanding of the associated cause-effect relationships within a decommissioning for nuclear facilities. The hazards associated with a decommissioning plan are important not only because they may be a direct cause of harm to workers but also because their occurrence may, indirectly, result in increased radiological and non-radiological hazards. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities as well as during accidents. Therefore, to prepare the safety assessment for decommissioning of nuclear facilities, the radiological and non-radiological hazards should be systematically identified and classified. With a semantic differential method of screening factor and risk perception factor, the radiological and non-radiological hazards are screened and identified.     

A study of the decommissioning procedure of an activated structure through an evaluation of the decommissioning cost for a research reactor

As decommissioning of a research reactor and a nuclear installation requires a long period of time from the decommissioning preparation work to the site remediation, the management of the data generated during the entire period of decommissioning is one of the most important tasks. In particular, the data obtained from research reactor decontamination and decommissioning activities can be important resources securing the safety and economic feasibility for other research reactor decommissioning. The owner of the research reactor and nuclear power plant need to submit decommissioning plan to the regulatory body at the starting stage of the research reactor and nuclear installation decommissioning project. The cost plan for decommissioning and the method for assessing the amount of exposure to protect workers must be stated in the decommissioning plan.This paper introduces the DES (Decommissioning Engineering System) that can be able to manage the data generated in the process of decommissioning of the TRIGA research reactor, to calculate an amount of waste, to evaluate decommissioning cost after deriving unit work productivity factor, and to predict the decommissioning process in advance. To verify the usability of this system and data integrity through connections among the unit systems, it describes the process to calculate the decommissioning cost using the data generated in dismantling an activated bio-shielding concrete in the TRIGA research reactor.As a result of the experiment to calculate the decommissioning cost with the TRIGA research reactor structure, it was found that the calculations were done precisely without flaw as the purpose of the experiment. Therefore, the DES can not only be used for other research reactors decommissioning, but also it is expected to be applied to other research reactors in the future. As a decommissioning cost between an activated concrete and a non-activated concrete according to the method of the dismantling procedure was significantly different, a study regarding the dismantling procedure needs more research.     

Development of Telerobotic Systems for Reactor Decommissioning, (III)

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program.     

Dismantling design for the loop rooms on the MR reactor

The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.     

石墨水冷型反应堆工程退役方案

本文简要介绍了反应堆工程退役的基本概念，石墨不冷堆的四大特点，石墨水冷堆退役必须遵循的基本原则，石墨水冷堆退役方案的编制和退役程序等。     

γ辐照装置退役环境影响评价要求

2011年环境保护部发布了《放射性同位素与射线装置安全和防护管理办法》,进一步明确了γ辐照装置退役环境影响评价的内容。本文介绍了γ辐照装置退役项目的环境影响评价要求,并就实践中的情况提出了几个建议。     

液态金属反应堆湿式燃料贮存桶内部事件概率安全分析

液态金属冷却剂在给反应堆带来运行安全与热效率优势的同时,也给反应堆带来了复杂的换料系统,其中大型液态金属反应堆采用的湿式乏燃料贮存桶是乏燃料卸料过程的核心设备,临时装载了大量的乏燃料组件,具备一定的安全风险。本文采用概率安全分析（PSA）方法对乏燃料贮存桶进行风险评价,通过运行状态分析、始发事件分析、事故序列分析以及简单的定量化,初步获得其导致乏燃料组件发生损伤的事故序列和最小割集,识别了关键系统与设备。结果表明,相对于反应堆本身的风险,乏燃料贮存桶本身风险虽低但依然不可忽略,且风险评价结果对反应堆的运行方式以及清洗系统的可靠性较为敏感。此外还对该系统的设计改进与安全优化进行了讨论。     

Preparations for decommissioning the MR research reactor at the Russian Science Center Kurchatov institute

The preparations of the MP research loop reactor for decommissioning are described. After final shutdown in 1993, a system of measures to ensure the required level of nuclear and radiation safety for the complex was implemented and a concept and fundamental program for decommissioning the reactor were developed. The variant DECON was taken as the base variant for decommissioning — immediate staged disassembly of structures and equipment, including disassembly of in-vessel structures of the RFT predecessor reactor which are stored in a central room of the reactor. The main results of the work performed to normalize the radiation conditions in the central room and to examine the safety-and-control system repository in this room are presented. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 259–264, May, 2008.     

A qualitative identification and analysis of hazards,risks and operating procedures for a decommissioning safety assessment of a nuclear research reactor

Korea Research Reactor-1(KRR-1, TRIGA Mark-II type reactor), the first nuclear research reactor in Korea, is being prepared for a decommissioning. The decommissioning methods and procedures of KRR-1 ought to be based on its structural conditions and radiological characteristics. Also, a systematic approach to the decommissioning tasks must be followed by reviews and assessments of the decommissioning workers’ safety.