福岛事故废液滞留和循环处理进程分析

日本福岛事故后,东京电力公司利用福岛第一核电厂原有的反应堆厂房、汽轮机厂房、高温焚烧厂房、工艺主厂房以及新设置的废液净化处理装置和废液贮槽,在现场建立了事故后废液滞留和循环处理系统,基本实现了事故废液的贮存、净化和循环利用。自2011年6月29日开始,东京电力公司约以周为间隔实时发布现场废液积存量/处理状况报告。本文统计了2011年6月29日—2014年1月8日期间先后发布的132份状况报告中有关现场各相关厂房和贮槽内废液积存量的数据,分析了福岛事故后废液滞留和循环处理进程,研究了福岛事故后建立的废液滞留和循环处理系统的组成、功能和运行特点及将其应用于废液滞留和循环处理中的经验与不足,并提出了关于事故后废液包容滞留的若干建议,为我国开展核电厂事故废液包容滞留措施研究提供借鉴。     

M310堆型及其改进型核电厂中压移动电源设置的有效性分析

参考岭澳核电站二期在移动电源方面的设计方案,使用概率安全分析（PSA）方法对《福岛核事故后核电厂改进行动通用技术要求（试行）》中中压移动电源设置的有效性进行分析。通过定性、定量分析,认为中压移动电源在功率和停堆工况下对全厂断电事故有较为明显的缓解效果,并对《福岛核事故后核电厂改进行动通用技术要求（试行）》提出了改进建议。      

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.     

浅议大亚湾和岭澳核电厂严重事故管理导则

介绍了大亚湾和岭澳核电厂严重事故管理导则,重点是大亚湾和岭澳核电厂严重事故管理导则的组成和特点,并对福岛核事故后大亚湾和岭澳核电厂所做的改进进行了介绍.     

日本311地震海啸导致福岛第一核电厂严重事故的三简

本文写于日本311大地震、海啸发生后的不同时段.首先,对福岛核事故的趋向进行了初步探讨,提出了发生堆芯熔化的可能性极大;其次,对我国的核安全检查进行了思考,述说了可能危及核电厂安全的外部自然事件和人为事件;最后结合地震后所发生的一些现象,提出了需要对地质、地震研究的问题.     

Large-scale seismic test program at Hualien Taiwan

The large scale seismic test (LSST) program at Hualien, Taiwan, is a follow-on to the soil-structure interaction (SSI) experiments at Lotung, Taiwan. The planned SSI studies are performed at a stiff soil site in Hualien, Taiwan, that historically has had slightly more destructive earthquakes in the past than Lotung. The objectives of the LSST program are as follows: to obtain earthquake-induced SSI data at a stiff soil site having similar prototypical nuclear power plant soil conditions; to confirm the findings and methodologies validated against the Lotung soft soil SSI data for prototypical plant condition applications; to validate further the technical basis of realistic SSI analysis approaches; to support further the resolution of USI A-40 “Seismic Design Criteria” issue. These objectives are accomplished through an integrated and carefully planned experimental program consisting of soil characterization, test model design and field construction, instrumentation layout and deployment, in situ geophysical information collection, forced vibration test, and synthesis of results and findings. The Hualien LSST is a joint effort among many interested parties. The Electric Power Research Institute (EPRI) and Taiwan Power Company (Taipower) are the organizers of the program and have the lead in planning and managing the program. Other organizations participating in the LSST program are the US Nuclear Regulatory Commission (NRC), Central Research Institute of Electric Power Industry (CRIEPI), Tokyo Electric Power Compamy (TEPCO), Commissariat A L'Energie Atomique (CEA), Electricité de France (EdF), Framatome, Korea Electric Power Corporation (KEPCO), Korea Institute of Nuclear Safety (KINS), and Korea Power Engineering Company (KOPEC). The LSST array started operation in June 1993, and is envisioned to be of five years duration.     

A simple mass and heat balance model for estimating plant conditions during the Fukushima Dai-ichi NPP accident

A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Dai-ichi nuclear power plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants had been limited by radiation, analytical investigation for the plant was required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named “HOTCB”, based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperatures up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi nuclear power plant (NPP) as examples to show the usefulness of the code.     

The erection and commissioning of the Mülheim-Kärlich Nuclear Power Plant

The contract for the Mülheim-Kärlich Nuclear Power Plant, equipped with a 1300 MWe pressurized water reactor, was awarded in 1973. The erection phase of the plant had been characterized in the mid-seventies by an aggravation of circumstances in connection with the nuclear energy controversy in the Federal Republic of Germany, the tightening of stipulations regarding safety philosophy, regulations and documentation, and by the consequences of the TMI accident in 1979. These led to considerable additional difficulties and delays. The commissioning phase on the other hand proceeded smoothly and speedily without major disturbances. The Mülheim-Kärlich Nuclear Power Plant has some major technical features distinguishing it from other pressurized water reactor plants built in the Federal Republic of Germany. Its nuclear steam system is based on a license from the Babcock & Wilcox Company, USA, but it was adapted to German rules and regulations. The Mülheim-Kärlich power plant is the first of this type and size built and put into operation. Its main technical features are described and, after a brief survey of the erection phase, the results of the start-up operations are discussed.     

核电站人因失效分析与预防

根据大亚湾核电站自身实践和世界同行的一般性经验 ,对核电站人因失效的特点、常见种类 ,以及产生人因失效的个人因素、外部因素和决策因素等主要因素进行了分析 ,有针对性地提出科学有效的预防措施 ,并重点介绍了大亚湾核电站在预防人因失效方面所取得的成功经验。     

Analysis of cesium isotope compositions in environmental samples by thermal ionization mass spectrometry-3

Cesium was recovered from soil samples obtained in Fukushima prefecture. Isotopic analysis of Cs was performed by γ-spectrometry to determine the activity ratio of ¹³⁴Cs/¹³⁷Cs and thermal ionization mass spectrometry was used to determine the isotopic ratios of ¹³³Cs/¹³⁷Cs and ¹³⁵Cs/¹³⁷Cs. The analytical results showed that both the activity ratio of ¹³⁴Cs/¹³⁷Cs and the isotopic ratio of ¹³⁵Cs/¹³⁷Cs were within the expected values for the Fukushima Daiichi Nuclear Power Plant estimated using the ORIGEN-II code, suggesting that most of the radioactive Cs in the soil sample originated from the Fukushima Daiichi Nuclear Power Plant. The concentration of ¹³⁷Cs and the contribution of radioactive Cs from global fallout were correlated to the distance from the Fukushima Daiichi Nuclear Power Plant, while the contribution of radioactive Cs from each reactor did not show any similar distance dependence.     

秦山第二核电厂场内医学应急救护计划

核电站发生核事故时的应急响应是一项十分复杂而庞大的系统工程,本文从医学应急救护的角度,参照《秦山第二核电厂场内应急计划》,并结合广东大亚湾核电站和秦山第一核电厂的经验反馈,探讨了秦山第二核电厂医学应急救护体系及其救护行动,并对做好场内医学应急救护准备工作提出了若干建议。     

Plant life management activities of LWR plants in Japan

Plant life management activities of Japanese LWR plants have been conducted since the early 1990s by the utilities and MITI (Ministry of International Trade and Industry) cooperatively. In Japan, where the regulatory practices are different from those in the US, there is neither law nor regulation that prescribes a licensed plant life for nuclear power plants. When an annual inspection is completed without any problem, the next cycle of operation would be permitted and this cycle can be repeated. However, it is generally known that mechanical components and structures deteriorate as they get older. So, we consider it very important to evaluate the long-term integrity of major systems, structures and components of old nuclear power plants. Japanese plant life management study consists of two parts. Both parts of the study were carried out confirming the integrity for the long-term operation of the three oldest Japanese LWR plants: Tsuruga Power Station Unit No.1 (BWR), Mihama Power Station Unit No.1 (PWR) and Fukushima Dai-ichi Nuclear Power Station Unit No.1 (BWR). The Part 1 study was conducted for the purpose of obtaining an outlook for long-term safety operation and was completed in 1996. The Part 2 study was conducted ensuring the plant integrity for the long-term operation in terms of, not only safety, but also reliability. The results of the Part 2 study were made public in February, 1999. Then, the recommended maintenance items were to be added to the existing maintenance programs of the three LWR plants.     

秦山核电厂核空气净化系统试验与运行

本文介绍了秦山核电厂核空气净化系统的设置,系统功能及运行方式,净化部件的性能参数。文章着重叙述安装工作完成后所进行的现场试验的项目、目的、方法和评价标准。已完成的现场试验证实,秦山核电站核空气净化系统将是可靠高效率的,文章最后介绍了运行管理的要点,指出强化运行管理是核空气净化系统能否发挥预定功能的关键。目前,秦山核电站正向满功率进军,核空气净化系统运行正常,满足了秦山核电厂安全运行的要求。     

核电站应急指挥中心基底隔震技术方案

日本福岛核电站事故之后,中国国家核安全局正式发布《福岛核事故后核电厂改进行动通用技术要求》,提高了核电站应急指挥中心的抗震安全要求。本文以防城港核电站应急指挥中心示范项目为分析对象,探讨了采用橡胶支座作为隔震装置的基底隔震方案,并进行了隔震效果评价。通过采用43个橡胶隔震支座,可以显著地削弱地震响应,提高应急指挥中心的抗震裕量。通过该技术方案可以保护应急指挥中心的人员、设备和系统的安全,从而提升核电站震后应急能力。分析研究成果和工程实践经验可在后续核电站建设中推广应用。     

Event sequence assessment of deep snow in sodium-cooled fast reactor based on continuous Markov chain Monte Carlo method with plant dynamics analysis

Margin assessment of a nuclear power plant against external hazards is one of the most important issues after Fukushima Dai-ichi Nuclear Power Plant Accident. In this paper, a new approach has been developed to assess the plant status during external hazards and countermeasures against them in operation quantitatively and stochastically. A continuous Markov chain Monte Carlo (CMMC) method is applied and coupled with a plant dynamics analysis. In the CMMC method, a subsequence plant status is determined by the latest state (Markov chain) and the status is evaluated from the plant dynamics analysis. A failure or success of safety function of plant component is also evaluated stochastically based on a latest state of plant or hazard. A numerical investigation of plant dynamics analysis against a snow hazard is also carried out in a loop type sodium-cooled fast reactor so as to assess the margin against the hazard.     

Modal parameter identification of a containment using ambient vibration measurements

Ensuring and maintaining the structural integrity of the containment structure in nuclear power plants is essential for preserving the nuclear reactor and other safety-related systems as well as protecting plant workers and publics from hazardous radioactive materials. To date, the structural integrity of the containment has been evaluated periodically via various nondestructive inspection methods. However, these methods require considerable time and cost to estimate overall structural integrity. In this paper, the possibility of monitoring the structural integrity of the containment utilizing ambient vibration measurement is explored. The ambient vibration testing was selected because it can avoid the interruption of normal operation of power plants. To fulfill the objective, the ambient vibration of the containment of Ulchin Nuclear Power Plant Unit 5 in Korea was measured, and the modal parameters, i.e., resonant frequencies and corresponding mode shapes, were extracted using the modal identification techniques in the frequency domain, i.e., the peak picking and the frequency domain decomposition methods. Using the extracted modal parameters and the finite element model, the elastic modulus of the concrete was estimated based on the sensitivity-based system identification method.     

Neutronic and thermo-hydraulic analyses of a small,long-life HTGR for passive decay-heat removal

Since the accident at Fukushima Daiichi Nuclear Power Plant in 2011, design concepts for nuclear reactors have been reconsidered with much greater emphasis placed upon passive systems for decay-heat removal. By considering this issue, the design parameter conditions for high temperature gas-cooled reactors (HTGRs) with passive safety features of decay-heat removal were obtained by residual-heat transfer calculation using equations for fundamental heat transfer mechanisms in our previous works. In the present study, the appropriate size of reactor core for a 100 MW_t reactor operating at 1123 K of the initial core temperature was found using the conditions. Consequently, neutronics and thermo-hydraulic analyses for the proposed reactor core were performed and the proper optimizations to control the excess reactivity and flatten the change in power peaking factor during operation were done successfully. By the systematic method to decide the core design which satisfies the condition for passive decay-heat removal, a long-life small HTGR concept whose excess reactivity was small during the operation was shown. The small excess reactivity is a significant advantage from the view point of safety in reactivity accident.     

Radiation imaging using a compact Compton camera mounted on a crawler robot inside reactor buildings of Fukushima Daiichi Nuclear Power Station

ABSTRACT

The Fukushima Daiichi Nuclear Power Station (FDNPS), operated by Tokyo Electric Power Company Holdings, Inc., went into meltdown in the aftermath of a large tsunami caused by the Great East Japan Earthquake of 11 March 2011. Measurement of radiation distribution inside the FDNPS buildings is indispensable to execute decommissioning tasks in the reactor buildings. We conducted a radiation imaging experiment inside the reactor building of Unit 1 of FDNPS by using a compact Compton camera mounted on a crawler robot and remotely visualized gamma-rays streaming from deep inside the reactor building. Moreover, we drew a radiation image obtained using the Compton camera onto the three-dimensional (3-D) structural model of the experimental environment created using photogrammetry. In addition, the 3-D model of the real working environment, including the radiation image, was imported into the virtual space of the virtual reality system. These visualization techniques help workers recognize radioactive contamination easily and decrease their own exposure to radiation because the contamination cannot be observed with the naked eye.     

从大亚湾三环路到秦山二期两环路看我国核电站自主设计能力

秦山核电二期工程与大亚湾核电站相比,反应堆功率下降,冷却剂系统三环路变为两环路。因此,重新设计了有关系统,并进行了相应的试验研究与工程验证。初步运行结果表明,设计是成功的。本文介绍部分系统的设计变化与特点以及相关的试验研究,反映设计与试验研究的成功结合,展现我国自主发展核电、自主设计大中型商用核电站的能力。