某核电厂放射性固体废物处理最小化实践

介绍了某核电厂在滤芯升级、在线去污、废物分类收集处理、多滤芯支架、大修废物产生量控制等一系列放射性固体废物最小化工作的实践经验;通过对该核电厂商运以来历年的放射性固体废物产生量进行分析,肯定了该核电厂在放射性固体废物最小化、废物管理工作方面取得的成绩。     

核电厂通风系统高效过滤器现场检测方法改进

荧光素钠法是核电厂通风系统高效过滤器效率现场检测方法之一, 在检测过程中部分系统为不可用状态。本文通过对检测方法进行分析并进行现场验证, 提出通过提高荧光素钠发生器的气溶胶质量流量和采样流量以缩短通风系统不可用状态持续时间的改进措施。     

国内核电厂水过滤器滤芯标准现状与讨论

介绍了核电厂水过滤器滤芯产品技术标准和性能检测标准及其现状,分析了核电厂水过滤器滤芯标准的发展趋势,指出了滤芯性能检测标准制定中亟待解决的难题,为核电厂水过滤器滤芯产品技术标准和性能检测标准的制定和完善奠定了基础。     

秦山核电基地放射性废物最小化技术实践与探讨

介绍了秦山核电基地在放射性废物最小化方面的实践。包括建立了蒸汽发生器排污树脂和通风过滤器金属框架清洁解控工作流程;改进了水泥固化线工艺,水泥固化包装容器采用金属桶代替水泥桶;开发了废过滤器芯子暂存衰变法分类处理及一桶装多芯方案;开发了技术废物“三明治”式废物装桶和超级压实提高外包装容器效率等最小化实践。对废液、废树脂、技术废物、有机废液等处理技术进行了比较分析,建议采用废液蒸干压实技术、废树脂蒸汽重整技术、可燃废物和有机废液焚烧技术,预计秦山核电基地废物产生量可由541 m³/a减至约94 m³/a。     

秦山三期重水堆核电站废过滤器滤芯转运与收贮

简述秦山三期重水堆核电站废过滤器滤芯的产生系统及产生量,废滤芯的转运的准备、实施,收贮入库等实施操作流程及其资料的整理。     

核电站低中放废树脂热态超压处理技术应用探讨

介绍一种从国外引进并首次应用于国内核电站的低中放废树脂有效减容处理技术--热态超压（超级压缩）处理技术,探讨了该技术在处理核电站低中放废树脂中的优势和今后需进一步关注的技术问题。     

核电厂运行准备与实践

核电厂的运行准备,是核电厂生产准备中的重要工作,由于核电厂从开工到投产的建设工期较长,一般需要5年左右,且核电运行的专业性要求很强,对核安全的要求也极高。特别是日本福岛事故后,国家层面、监管部门和各大核电集团都对核电厂的安全运行提出了更高要求。依据核电机组建设阶段的特点及结合工程实践,文章阐述了核电厂运行领域的体系建立,并概括性分析了相关管理创新。     

秦山第二核电厂主环路温度计故障分析与处理

针对秦山第二核电厂1、2号机组运行初期发生的主环路温度计的典型故障,通过对温度计进行解剖和材料分析、与其他核电厂对比分析、试验室和现场振动分析,确定主环路温度计的故障原因,提出处理方案。结果表明,温度计失效原因为套管的强度不够大,套管与温度计发生碰撞导致温度计末端损坏和破裂;温度计外壁加环改进方案可有效解决故障问题。     

后处理设施事故分类方法和验收准则探讨

为推动我国后处理设施事故分类方法和验收准则进一步发展完善,对比研究了美国历史工程经验和美国当前监管要求。研究结果表明,美国当前监管要求在历史工程经验采用的确定论方法基础上引入了风险评价。通过对设施的潜在危害进行梳理和风险分析,识别出需要关注的危害项;通过对事故的可能性和后果进行风险分析,以风险边界作为事故验收依据。所选取的事故分析边界既包括历史工程经验常用的设计基准事故,也包括后果轻微但发生可能性大的小事故。美国当前监管要求评价了工程措施的有效性,论证了设施安全设计的合理性,具有重要的借鉴或参考价值。     

球床高温气冷堆核电站放射性废物最小化策略研究

放射性废物最小化是放射性废物管理的基本原则之一。高温气冷堆核电站放射性废物最小化策略研究对于优化设计与运行实践和全寿期放射性废物管理,以及高温气冷堆产业化发展具有重要意义。通过对世界上主要球床高温气冷堆运行历史和放射性废物数据的调研和论证,分析了球床高温气冷堆技术及其放射性废物特点,总结了高温气冷堆放射性废物管理值得借鉴经验及相关研究进展,提出了高温气冷堆核电站全寿期放射性废物最小化的策略和建议。     

Computational and experimental validation of protective features of transport packages for long-term storage of spent nuclear fuel from nuclear power plants

Abstract

The purpose of this paper is to describe the work of our Institute (RFNC-VNIITF) on the development of a unique long transport container using new manufacturing technologies. The container is designed for transport and long-term storage of spent fuel elements more than 10 m long. The disposal of such elements usually implies their cutting. This requires considerable expense in building special cutting plants. Another way is to use long protective shells to contain and store the entire fuel cell. Development of such containers has been made possible by the use of rolled-strip technology ensuring maximum strength of containers. The paper contains certain results of calculations and experimental research justifying and confirming the protective properties of the long package. To confirm reliability with minimum material consumption an available package model, manufactured by new technology, is used. This package was enhanced in the flange area to simulate strictly a flange joint of a full-scale transport package. Calculations and experimental work were performed according to IAEA regulations. During investigations the full-scale cask was considered as both a model and a prototype as it simulated completely the bottom and the flange joint 'lid-cask body' (i.e. being a model) while being shorter and thinner than the cylindrical part of the body (i.e. being a prototype). Here it is called simply a prototype. The complete simulation was not set as an objective. At the same time the cask body loading for tests simulating an air crash was done with regard to the full-scale design. The paper contains results obtained analytically on the basis of experimental data to justify the properties of a full-scale package without experimentally testing it: therefore only some photographs and experimental results for the prototype are presented here. The title of the paper indicates that the analytical study was conducted with use of numerical constants obtained experimentally. It is impossible to present all the information, including experimental research, here because of the large scope of the work performed and the limited space available.     

海阳核电一期工程建设管理与实践

介绍了山东海阳核电一期工程的基本情况,以及项目管理和工程进展情况。通过海阳核电工程建设管理实践,提出了核电建设中遇到的问题,供同行进行交流和探讨。     

Heat-treatment and storage technology for spent ion-exchange resins for new-generation nuclear power plants

A technology for thermovacuum drying of spent radioactively contaminated ion-exchange resins was developed. It was shown experimentally that thermovacuum drying yields a product suitable for long-term storage and subsequent disposal. Recommendations were made for using NZK and KMZ concrete and metal casks for holding dry, spent, low- and medium-activity ion-exchange resins without additional solidification. Because the volume of the product obtained is greatly reduced, thermovacuum drying will decrease by at least 15-fold (from 100 to 7) the number of casks required for long-term storage and subsequent disposal of spent ion-exchange resins produced in one year of operation of a single VVER-1200 power-generating unit.     

New proposal of reactivity coefficient estimation method using a gray-box model in nuclear power plants

A new method for estimating reactivity parameters, such as moderator temperature coefficient (MTC) and void reactivity coefficient (VRC), is proposed using steady-state noise data. In order to solve the ill-posed problem of reactivity parameter estimation, a concept of a gray box model is newly introduced. The gray box model includes a first principle based model and a black-box fitting model. The former model acts as a priori knowledge based constraints in a parameter estimation problem. After establishing the gray box and noise source models, the maximum likelihood estimation method based on Kalman filter is applied. Furthermore, it is shown that the frequency domain approach of the gray box model is useful in the case of VRC estimation. The effectiveness of the proposed algorithms is shown through numerical simulation and actual plant data analysis.     

Choice of organic diluents for the extractive regeneration of the spent fuel of nuclear power plants

Conclusions The length of the n-alkane chain which provides a basis for the diluent can lie within the interval C₁₁-C₁₅. Within this interval, the ratio of the separate n-alkanes in the diluent can be regulated. It is better, however, to use the hydrocarbons C₁₁-C₁₅; their rather high flash point gives them a slight edge over the higher members of the group in regard to their compatibility with the extracting solvates of the actinide nitrates and also with regard to their hydrodynamic characteristics. The content of aliphatic acids and alcohols 0.01 M; of unsaturated compounds 0.005 M; of aromatic hydrocarbons 1 vol.%. These requirements may change in the future as more investigations are made and a deeper study is made of the factors which affect the behavior of the diluent in the extractive cycle.Translated from Atomnaya Énergiya, Vol. 47, No. 2, pp. 75–79, August, 1979.     

核电工程管理程序评估创新与实践

介绍了山东核电有限公司在AP1000核电工程程序管理工作中创新采用的程序评估方法具体的实施方式和过程,总结了程序评估实施的效果,分析了程序评估这一管理方法具有的优势和存在的不足。     

球形自动反冲洗管道过滤器在核电站海水冷却系统中的应用

以田湾核电站1、2号机组为例,介绍了核电站海水冷却系统的组成、功能以及海水夹带杂物引起海水冷却系统换热器堵塞的问题;针对换热器堵塞问题,讨论了在核电站海水冷却系统安装自动反冲洗管道过滤器的必要性,介绍了真空泵海水冷却系统安装自动反冲洗过滤器设备选型、方案设计的一些特点,以及球形自动反冲洗过滤器在田湾核电站的应用效果。