铀矿地质分析测试技术回顾与新形势下网络实验室构建

回顾了目前铀矿地质分析测试所使用的元素分析技术、核素和同位素分析技术、元素形态分析技术、微区分析技术、流体包体与有机成分分析技术。提出了铀矿地质分析测试技术的几个重点发展方向,包括高精度质谱分析技术研究、专用标准物质研究、现场分析技术研究、微区原位分析技术研究和有机成分分析技术研究。探讨了新时期核工业地质分析测试网络实验室运作模式。主要思路是以核工业北京地质研究院分析测试研究中心为骨干,以核工业地质局各区域研究所和大队实验室为成员,建立更密切的协作关系,构建核工业地质分析网络实验室。实现资源共享,发挥其整体优势,更好地为核工业地质服务。     

船用核动力装置辐射安全技术研究

辐射安全技术是船用核动力装置辐射安全水平的根本保障,目前形势下我国船用核动力装置的辐射安全水平亟待提高。对辐射安全技术中处于核心地位的放射性源项分析技术和辐射屏蔽设计技术开展了国内外研究现状和趋势的调研分析,并从发展需求角度,归纳了两者的总体发展目标,梳理了各自的重难点技术,最后绘制了放射性源项分析技术和辐射屏蔽设计技术的技术发展路线图,为我国船用核动力辐射安全技术发展提供了方向和支撑。      

ICP-MS用于核保障安全监督环境样品超痕量核材料分析技术的开发

为环境样品超痕量核材料分析开始筹建洁净实验室并开发分析技术,包括整体分析和粒子分析。对于整体分析,采用了ICP-MS和TIMS技术进行U和/或Pu的浓度和同位素比值分析。而对于粒子分析,将使用SIMS测定每个粒子中的U和/或Pu的同位素比值。本文简要地介绍了分析技术的开发概貌并描述了目前使用ICP-MS进行整体分析的开发状况。     

与地浸砂岩型铀矿有关的分析测试技术发展趋势

在分析铀矿地质对分析测试需求的基础上,提出了今后几年与地浸砂岩型铀矿有关的分析测试技术的发展趋势。主要发展的技术是:以XRF和ICP-MS为主要手段的高精度、高灵敏度和高效率的元素分析技术;以多道α能谱、低本底高纯锗γ能谱和高精度质谱分析技术为主要手段的核素和同位素分析技术;以联用技术为主的元素形态分析技术;以荧光显微观测、电子探针和激光探针为主要手段的微区分析技术;以各种色谱技术为主的有机成分分析技术;以化学分析技术为主的微生物成分分析技术。此外,引入物理化学概念,解释砂岩铀矿成矿过程,有利于拓宽砂岩铀矿找矿思路,帮助砂岩铀矿地质分析参数的设置。     

井间示踪测试技术新进展

对油田井间示踪测试技术的发展概况进行简要叙述。分析了各代技术的优缺点;根据油藏动静态描述手段和原理及测试解释技术理论并结合矿场实践进展,对解释方法进行了对比分析,提出了从示踪剂选择到测试解释的优化技术体系;另外,示踪剂技术目前出现了多个分支,尚处于矿场试验阶段。本文给出了部分有关测试成果和解释成果,以及相关的分析和评价。井间示踪测试技术是目前成熟和先进的油藏开发动态测试技术之一,对于解决油田开发过程中面临的实际问题具有重要的实用价值和指导意义,应用前景广阔。     

基于核磁共振的癌症非侵入式诊断理论与技术研究

分析了核磁共振技术在癌症临床诊断领域的应用现状、相关理论与技术的发展趋势,概述了对核磁共振设备产生的FID信号、MRI和MRS的数据分析技术,并以MRS为重点,研究了FID信号和波谱数据的特点,提出了利用小波变换和模式识别技术对FID信号和波谱数据进行分析的技术方案,并对信号的小波阈值去噪、小波基函数的选择、MRS特征识别与提取等问题进行了探索性研究,对核磁共振为基础的相关技术在癌症临床诊断领域内应用的进一步发展提示了方向。     

虚拟装配技术及其在中国快堆技术中的应用展望

阐述虚拟装配技术的主要研究内容和现状,指出该技术的关键内容并分别进行分析.针对中国快堆技术发展中的机械设备装配现状,分析虚拟装配技术在快堆技术中应用的重要意义,提出了该技术在快堆技术应用中的发展目标.     

池式快堆三维热工流体力学设计研究

随着计算机软硬件技术的发展,三维数值分析技术已经成为池式快堆堆芯和钠池热工设计和计算分析的重要组成部分,并在其中发挥着不可替代的作用.通过对池式快堆几个典型热工现象的分析,展示了我国第一座池式快堆(中国实验快堆)热工设计和安全分析中所拥有的设计手段和工具,总结了三维数值分析技术在快堆工程中的应用,并指出了其对今后快堆热工设计的重要意义.     

池式钠冷快堆非能动停堆技术方案研究

非能动停堆技术一直是快堆安全技术研究的热点,受到了国际上研究快堆技术国家的重视。目前,国内也开展了相关技术研究。本文在调研和分析的基础上,从非能动停堆技术的安全特性、技术成熟度和中国实验快堆(CEFR)典型事故的分析等几个方面进行了两种有一定试验经验的非能动停堆装置的对比,并给出了中国池式钠冷快堆非能动停堆技术发展的建议。     

核材料与核设施实物保护延迟技术

分析了核材料与核设施实物保护延迟技术和延迟设施的现状,提出了延迟技术的改进措施和发展趋势.     

The IAEA accident management programme

Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff.     

核电厂事故条件下主控室可居留性剂量评价方法

基于美国核管会(NRC)对核电厂主控室可居留性评价的技术要求,并结合我国二代改进型核电厂的设计特点,提出完整的主控室内可居留性人员剂量评价模型。相对于国内现有的计算方法,新模型可兼顾设计基准事故与严重事故情况,采用组合尾流模式计算短期大气扩散因子,结合可居留区域划分及通风系统的设计特点考虑建筑物及通风系统的未经过滤的泄漏对评价计算的影响。     

福岛核事故后对我国核电厂应急控制中心抗震设防的思考

结合福岛核事故后对我国核电厂进行的核安全检查,分析了我国核安全法规关于核电厂应急控制中心的要求以及福岛核事故的经验教训,提出目前我国核电厂应急控制中心采用民用抗震设防标准进行抗震设防,无法保证在由地震引发的应急事故工况下应急控制中心的功能,应该适当提高其抗震设防级别。     

浅析中美核电站环境监测的通用要求

分析了中美两国核电站环境监测的管理机制及监测方案的通用技术要求,针对当前我国核电站环境监测现状提出了相关建议,旨在借鉴国外先进经验,促进我国核电站环境监测管理水平的提高。     

Seismic qualification of systems, structures, equipment and components

The purpose of this paper is to give an overview of the various qualification procedures available to the vendors of nuclear power plants and equipment for hopefully achieving NRC (Nuclear Regulatory Commission) plant licensing and overall guaranteed safe operation. These procedures usually involve computer-aided analyses for large systems and structures, but trend toward shaking table tests for small equipment and components.The dynamic analysis and testing required for seismic qualification can be covered in a practical manner by reference to several pertinent Regulatory Guides and Standards. They have been issued by the NRC on specific subjects, but often represent a consensus of more general standards prepared by ASME, IEEE, ASCE, ANSI and NEMA. These documents cover such diverse subjects as (a) reactor site criteria, (b) seismic design limits and loading combinations, (c) system damping values, and (d) recommended vibration test practices.The author has been directly concerned with IEEE Std 344 on seismic qualification practices and has therefore included the latest ideas and suggestions for revising this document. In general, there has been a continuing escalation in the g-level of seismic requirements. This present overview indicates a need for R&D work and re-examination of published documents to counterbalance unwarranted conservatism.     

Seismic analysis of nuclear power plants by using three-dimensional finite element models: a review

ABSTRACT

Seismic design of nuclear power plants (NPPs) is important for ensuring their integrity during earthquakes. Seismic analysis has been conducted using lumped mass beam models (LMBMs) for the design of plants in Japan, whereas three-dimensional (3D) finite element models (FEMs) have been used for novel plants outside Japan. The purposes of this study are to organize issues related to the development and application of 3D FEMs for seismic analysis of Japanese NPPs and to indicate future study directions. To organize these issues, the authors systematically investigated: (1) international guides and standards related to seismic analysis and (2) 3D FEMs of novel NPPs outside Japan. By considering other studies on the issues, the authors suggest directions for future studies. Resolving the issues will contribute to application of 3D FEMs for seismic analysis in the design of Japanese NPPs.     

运行核电厂抗震裕度评价研究

抗震裕度评价是对核电厂应对超过设计基准地震能力的评价,特别是日本福岛核事故发生后,评价核电厂应对超过设计基准外部事件时的安全裕量、优化和落实改进措施、提高改进措施的有效性就显得尤为重要。本文通过研究国际上广泛采用的抗震裕度评价方法,最终选定EPRI SMA方法对秦山第二核电厂进行抗震裕量分析。分析结果表明：秦山第二核电厂满足1.5倍SSE的抗震裕度要求,具有较强的抗震能力。     

内陆核电厂发展状况与安全监管探讨

对国际上内陆核电厂的发展状况进行了分析,详细介绍了我国内陆核电厂厂址的特点和评价情况,进而对国内首批内陆核电项目的审查情况和监管要求进行了全面的论述。在总结沿海核电厂建设实践和审评经验的基础上,结合我国发展内陆核电厂的实际情况,研究并提出了建设内陆核电厂的审评重点和监管要求。     

Simplified seismic analysis methods used by AECL for the seismic qualification of CANDU nuclear power plants

This paper presents some of the simplified procedures and methods used by AECL for the seismic qualification of CANDU Nuclear Power Plants (NPPs). The approaches described herein are well tested and have been used in Canada and elsewhere for a number of years. Most of these simplified seismic analysis, testing and inspection procedures, and their underlying principals, have been accepted by the Atomic Energy Control Board of Canada for licensing purposes. In this respect, a comprehensive inspection of completed NPPs, to determine their ability to safely survive a design basis earthquake (DBE), is a prerequisite for licensing of CANDU NPP's in Canada. Many of the methods and recommendations given in the following tie in closely with [1].     

A probabilistic seismic risk assessment procedure for nuclear power plants: (I) Methodology

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology.