日企提交有关后处理厂问题的最终报告

<正>【日本原子能产业协会网站2010年8月9日报道】2010年7月28日,日本核燃料公司(JNFL)向日本原子能安全保安院(NISA)提交了一份有关六所村     

日本原子能安全研究概况

根据日本科学技术厅编著的原子能手册(昭和58年版)提供的资料,现将有关日本原子能安全研究计划概况摘录如下。 (一) 日本原子能开发利用的基本方针根据民主、自主、公开的原则,确定原子能开发利用的如下方针。 1.坚持和平利用,批准核武器不扩散条约。2.确保安全,1978年原子能基本法的基本方针就是确保安全宗旨,为此,设立原子能安全委员会,实行安全规则的行政常规     

日本设立一个日本南朝鲜联络委员会

【日本《原子力产业新闻》1984年9月6日报道】日本原子能产业会议(原产)8月31日在东京召开了日本与南朝鲜原子能联络委员会的首次会议。日本与南朝鲜在原子能工业领域方面的合作,虽然过去从未间断过,但为了今后应     

美国帮助巴基斯坦保障核材料

【日本原子能研究所网站2004年2月9日报道】 日本原子能研究所在高崎研究所设立的作为国际监测系统(IMS)下属的放射性核素监测站于2004年2月6日获得了全面禁止核试验条约组织(CTBTO)筹备委员会的认证。原研自设立该监测站以来,每天获取数据并向位于维也纳的国际数据中心(IDC)传送。这次,观测所完成了由CTBTO进行的评估并通过了认证。IMS由采用地震波、放射性核素、水下声波、微气压振动四项监测技术的位于世界各地321处监测站和16处政府许可的测试设施构成。IMS将测定结果通过世界信息传送平台(GCI)发送到IDC,再由IDC向各国的…     

原子能文献库在日本开始联机服务

【日本《原子能快报》1984年1月9日第11页报道】日本科学技术情报中心(JICST)于1月5日在东京举行了“国际核情报系统原子能文献库”开始服务的开幕式。日本科学技术情报中心从1976年开始进行情报联机检索服务(JOIS)。这次,在原     

日本颁布两部有关原子能的白皮书

<正>【本刊2010年4月综合报道】最近,日本原子能委员会(AEC)与日本核安全委员会(NSC)分别出台了有关核能和核安全的白皮书,即2009年版《原子能白皮书》与2009年版《原子能安全白皮书》。     

日本原子能产业会议

简史日本原子能产业会议(JAIF)于1956年3月1日成立,是日本惟一的综合性民间原子能组织。它是日本原子能行业赞助的非盈利组织,旨在促进日本核能的和平利用,同时兼顾放射性同位素核辐射技术的和平利用。日本原子能产业会议的目标是:全面开展核能研究,为政府制定核能开发和利用计划提供支持,推进政府政策的实施,以及促进国家经济和福利的健康发展。组织机构日本原子能产业会议目前的理事长为西泽润一,副理事长5名、常任理事24名、理事58名、常任监事2名、监事2名。JAIF拥有许多日本国内外的会员,总数约100家。例如,与原子能开发关系密切的…     

日本公布《面向增长的原子能战略》

<正>【日本原子能委员会网站2010年4月27日报道】日本原子能委员会(AEC)于2010年4月27日公布了《面向增长的原子能战略》草案。该战略旨在切实推进原子能的发     

日本着手制造ITER超导线圈导体

<正>【日本原子能研究开发机构网站2010年1月22日报道】日本原子能研究开发机构(JAEA)与新日铁工程株式会社已根据国际热核聚变实验堆(ITER)超导线圈导体制造协议开展了多项相关筹备工作。最近,双     

来自日本的消息

1985年4月至1987年3月期间,日本保健物理学会(JHPS)的负责人员、秘书及办公处地址如下: 会长:Dr.Seiji Fukuda(东京,动力堆及核开发公司,保健及安全部) 副会长:Dr.Yoshikazu Yoshia(东海,日本原子能研究所,保健物理研究室) 秘书:Dr.Jun Akaishi(东海,日本原子能研究所,保健物理研究室)     

Review of five investigation committees’ reports on the Fukushima Dai-ichi nuclear power plant severe accident: focusing on accident progression and causes1

On March 11, 2011, the Tohoku District-off the Pacific Ocean Earthquake and the subsequent tsunami resulted in the severe core damage at TEPCO's Fukushima Dai-ichi Nuclear Power Plant Units 1–3, involving hydrogen explosions at Units 1, 3, and 4 and the large release of radioactive materials to the environment. Four independent committees were established by the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and TEPCO to investigate the accident and published their respective reports. Also, the Nuclear and Industrial Safety Agency carried out an analysis of accident causes to obtain the lessons learned from the accident and made its report public. This article reviews the reports and clarifies the differences in their positions, from the technological point of view, focusing on the accident progression and causes. Moreover, the undiscussed issues are identified to provide insights useful for the near-term regulatory activities including accident investigation by the Nuclear Regulation Authority.     

Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident

Research and development in nuclear reactor physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, universities, and corporations, focusing on the areas in reactor physics and thermal-hydraulics since the Fukushima accident to the present date.     

柏崎·刈羽核电厂地震安全评价简介

介绍了日本柏崎·刈羽核电厂遭受强烈地震时的情况,以及国际原子能机构（IAEA）专家评审组在现场发现的问题及其经验反馈。     

HAF003修订中一些问题的探讨

基于对国内外核电质量保证实践的对比分析与研究,论述了在《核电厂质量保证安全规定》(HAF003)的修订过程中需要考虑和关注的事项,涉及管理职责、独立评价、设计控制和采购控制等24个问题,以期为国家核安全局主持的HAF003修订工作提供参考和建议。     

Review of studies on criticality safety evaluation and criticality experiment methods

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS.     

A code for simulation of human failure events in nuclear power plants: SIMPROC

Over the past years, many Nuclear Power Plant organizations have performed Probabilistic Safety Assessments to identify and understand key plant vulnerabilities. As part of enhancing the PSA quality, the Human Reliability Analysis is essential to make a realistic evaluation of safety and about the potential facility's weaknesses. Moreover, it has to be noted that HRA continues to be a large source of uncertainty in the PSAs. Within their current joint collaborative activities, Indizen, Universidad Politécnica de Madrid and Consejo de Seguridad Nuclear have developed the so-called SIMulator of PROCedures (SIMPROC), a tool aiming at simulate events related with human actions and able to interact with a plant simulation model. The tool helps the analyst to quantify the importance of human actions in the final plant state. Among others, the main goal of SIMPROC is to check the Emergency Operating Procedures being used by operating crew in order to lead the plant to a safe shutdown plant state. Currently SIMPROC is coupled with the SCAIS software package (Izquierdo et al., 2008), but the tool is flexible enough to be linked to other plant simulation codes.SIMPROC-SCAIS applications are shown in the present article to illustrate the tool performance. The applications were developed in the framework of the Nuclear Energy Agency project on Safety Margin Assessment and Applications (SM2A).First an introductory example was performed to obtain the damage domain boundary of a selected sequence from a SBLOCA. Secondly, the damage domain area of a selected sequence from a loss of Component Cooling Water with a subsequent seal LOCA was calculated. SIMPROC simulates the corresponding human actions in both cases.The results achieved shown how the system can be adapted to a wide range of purposes such as Dynamic Event Tree delineation, Emergency Operating Procedures and damage domain search.     

Integrating Safety Critical Software System in Probabilistic Safety Assessment

In the recent times, computer-based systems are frequently used for protection and control of Nuclear Power Plants (NPPs). In the conventional Probabilistic Safety Assessment (PSA), the contribution from software in these computer-based systems was not given necessary attention. However, from operating experience, it has been found failures in such systems can also result in initiating events that have the potential for leading into Core Damage in the event of unavailability of the respective Engineered Safety Features. The impact of a typical computer-based system on PSA of Indian Nuclear Power Plant is demonstrated.     

Normalization of discharges and emissions of radionuclides and harmful substances into the environment

All-Union Scientific-Research Institute of Nuclear Power Plants. Odessa Polytechnical University. Chernobyl Nuclear Power Plant. Beloyarsk Nuclear Power Plant. Ministry of the Environment, Ukraine. TSMSCH-5, Ukraine. Translated from Atomnaya énergiya, Vol. 76, No. 6, pp. 522–526, June, 1994.     

Development of a short-term emergency assessment system of the marine environmental radioactivity around Japan*

The Japan Atomic Energy Agency (JAEA) has, for many years, been developing a radionuclide dispersion model for the ocean, and has validated the model through application in many sea areas using oceanic flow fields calculated by the oceanic circulation model. The Fukushima Dai-ichi Nuclear Power Station accident caused marine pollution by artificial radioactive materials to the North Pacific, especially to coastal waters northeast of mainland Japan. In order to investigate the migration of radionuclides in the ocean caused by this severe accident, studies using marine dispersion simulations have been carried out by JAEA. Based on these as well as the previous studies, JAEA has developed the Short-Term Emergency Assessment system of Marine Environmental Radioactivity (STEAMER) to immediately predict the radionuclide concentration around Japan in case of a nuclear accident. Coupling the STEAMER with the emergency atmospheric dispersion prediction system, such as Worldwide version of System for Prediction of Environmental Emergency Dose Information version II (WSPEEDI-II), enables comprehensive environmental pollution prediction both in air and the ocean.     

国际原子能机构SSG-30安全分级方法应用解析

安全分级的目的是确保物项的设计、制造、建造、调试和运行采用恰当的要求,使物项在所有预期的运行工况下有适宜的质量,进而确保安全功能的实现。国际原子能机构（IAEA）2014年颁布的核电厂构筑物、系统和部件（SSC）安全分级导则（SSG-30）,其安全分级原则涵盖核电厂5个纵深防御层次,从设计预防措施和安全功能分类两个维度识别安全相关物项的重要性,考虑核电厂运行工况状态和放射性与运行限值的要求,进而确定物项的安全级别和相关的规范要求。