Japan Atomic Power Company Research Articles

Elimination of recriticality potential for the Self-Consistent Nuclear Energy System

The ultimate safety goal of the Self-consistent Nuclear Energy System (SCNES) is to eliminate the recriticality-problem based on a simple safety logic. The principle of the elimination of the recriticality-problem is the Controlled Material Relocation (CMR) to establish the neutronic shutdown by removing the molten fuel to the out of core before a large scale pool formation which has potential of energetics driven by a super prompt criticality. The CMR concept should be reliable without significant impact on the core neutronic performance. As the typical core concepts to enhance this CMR characteristic, several design options are under consideration. They are fuel assemblies with inner duct structure (FAIDUS), fuel assemblies with hollow fuel pins in the axial blanket region (ABLE) for MOX fueled cores, and fuel assemblies without fuel pin bundle structure in the lower axial blanket region (ELAB) for the metallic fueled core. Based on the core design study and accident analyses, these CMR-oriented concepts have been found feasible without significant degradation of the neutronic performance In order to experimentally confirm the effectiveness of the CMR concept for the MOX fueled core, the EAGLE project has been started in 1998 by Japan Nuclear Cycle Development Institute (JNC) and The Japan Atomic Power Company (JAPC). The EAGLE project is the experimental program utilizing the out-of pile test facility and in-pile facility IGR of the National Nuclear Center of the Republic of Kazakhstan (NNC/RK).

P21-01 革新的中小型炉について

For a solution to the problems, such as poor natural resources and global warming, nuclear power should be considered as a future power source. In the situations which prefer a plant siting near a consumption area to reduce large-scale power failure risk and electricity power transmission costs and which require low-risk investments under the liberalization of retail market for electricity, small-medium sized reactors (SMRs) may be more useful than large ones. At one time it had been considered that the SMR was less economical than a large reactor for its scale-demerit. Recently the SMR have been enhanced its economic competitiveness by introducing innovative technologies. So The Japan Atomic Power Company (JAPC), together with the Central Research Institute of Electric Power Industries (CRIEPI) started on September 2000 a preliminary investigation to assess the potentialities of SMRs and the significance of them as a future energy option. From the investigation, it has been confirmed that the SMRs with the innovative technologies have several advantages, such as smaller financial risk than large plant, more easily understandable safety than conventional plant. And the mass production effect and the series construction effect being considered, the SMR's power generation costs will be almost equivalent to that of a large plant. So the investigation have come to the conclusion that the SMRs with innovative technologies are a promising energy option for the future.

技術報告 ３４２． 原子炉施設の廃止措置計画策定および管理のための計算システムの開発 東海発電所の解体作業に関するプロジェクト管理データの試算

Data on the Japan Power Demonstration Reactor dismantling activities were analyzed to be applied to planning and evaluation of decommissioning reactor facilities. These were then combined with the computer systems for planning and management of reactor decommissioning (COSMARD) as a database to estimate decommissioning project management data such as manpower needs and radiation exposure to workers. The Tokai Power Station decommissioning program, which will be conducted by the Japan Atomic Power Company as the first project on decommissioning a commercial nuclear power plant in Japan, was characterized by evaluating the project management data using COSMARD and the database. It was found from the study that COSMARD and unit productivity factors were applicable to planning and evaluation of reactor decommissioning programs.

Elevated-Temperature Mechanical Properties of an Advanced-Type 316 Stainless Steel1

Type 316FR stainless steel is a candidate material for the Japanese demonstration fast breeder reactor plant to be built in Japan early in the next century. Like type 316L(N), it is a low-carbon grade of stainless steel with a more closely specified nitrogen content and chemistry optimized to enhance elevated-temperature performance. Early in 1994, under sponsorship of The Japan Atomic Power Company, work was initiated at Oak Ridge National Laboratory (ORNL) aimed at obtaining an elevated-temperature mechanical-properties database on a single heat of this material. The product form was 50-mm plate manufactured by the Nippon Steel Corporation. Data include results from long-term creep-rupture tests conducted at temperatures of 500 to 600°C with test times up to nearly 40.000 h, continuous-cycle strain-controlled fatigue test results over the same temperature range, limited creep-fatigue data at 550 and 600°C, and tensile test properties from room temperature to 650°C. The ORNL data were compared with data obtained from several different heats and product forms of this material obtained at Japanese laboratories. The data were also compared with results from predictive equations developed for this material and with data available for types 316 and 316L(N) stainless steel.

Design study of the demonstration FBR in Japan

The objective of the plant design study Phase 2, conducted by the Japan Atomic Power Company since 1997 for 3 years, is to accomplished a plant overall concept of the Demonstrative FBR (DFBR) that has economical potential toward commercialization and offers high reliability to plant operators not to cause a long unexpected shutdown resulting from a trouble, i.e., sodium leakage or fires. This has been successfully achieved by establishment of a plant overall design of 672 MWe consisting of the reactor system with drastically simplified internals, the compact and double walled coolant boundaries, the well rationalized fuel handling system, the BOP systems introducing up-to-date LWR equipment, and the compact reactor building. The plant construction cost has been estimated based on the quantity of materials to be about 130 % on the bases of a 1000 MW LWR, which is well contented with the requirement. The DFBR plant concept, having economical potential toward commercialization, safety and reliability, has been established in the plant design study Phase 2.

The present status of DFBR development in Japan

The first demonstration fast breeder reactor (DFBR-1) is being developed in Japan, as the essential step toward the commercial deployment of FBR around 2030. The construction of the DFBR-1, which follows the experimental FBR (JOYO) and the prototype FBR (MONJU), is scheduled for the early 2000s. The Japan Atomic Power Company (JAPC) has been conducting design studies of the DFBR-1 and relevant R&D since 1986, under a contract awarded by the nine electric companies and the Electric Power Development Co., Ltd. The electric utilities have settled the major design specifications of the DFBR-1 plant as the top-entry loop type system with an output of 660MWe. Based on these major design specifications, JAPC has been conducting the optimization design study of the DFBR-1 since FY 1994 in order to establish the overall consistent plant design with construction cost reduction below 1.5 times that of LWRs on 1,000 MWe basis and comfirm the licensability of the DFBR-1 design. This study will complete by the end of FY 1996 This paper mainly outlines the design concepts of DFBR-1 resulting from the optimization design study of DFBR-1.

東海発電所の廃止措置

日本初の商業用原子力発電所である東海発電所は, 1998年3月末をもって運転を終了した。その建設,運転を通して我が国の原子力発電の基礎を築き,多くの原子力関係者を育成した。 32年間にわたり運転経験を積み重ね,原子力発電の定着という役割を果たしてきた。運転の終了は同時に廃止措置の始まりでもあり,東海発電所は日本では未だ経験のない商業用原子力発電所の安全かつ合理的な廃止措置の実証という新たな役割を担った。

Development of the demonstration FBR and its commercialization

The FBR being developed in Japan is expected to operate commercially by the year 2030 or so. The construction of the first demonstration FBR “DFBR-1”, which follows the experimental FBR “JOYO” and the prototype FBR “MONJU” is scheduled for early 2000s. The Atomic Energy Commission adopted these policies in the new version of “The Long Term Program for Research, Development and Utilization of Nuclear Energy”. The Japan Atomic Power Company has progressed design studies of the DFBR-1 and R&Ds. Electric utilities settled the major design specifications of the DFBR-1 such as the top entry loop type as the reactor type and about 660 MWe of plant capacity. To achieve the goals for the commercialization of FBRs, such as a high level of safety and reliability comparable with a LWR and further improvement of economy, many innovative technologies and system improvements will be required in commercial FBRs. This paper describes the present status of development of the DFBR-1.

Development, operational experience and implications for future design of FBRS in Japan

Joyo, the 100 MW t experimental reactor, has been successfully operated since 1977, and Monju, the 280 MW e prototype FBR, is under construction, with the first criticality planned for 1992. To promote FBR R&D efficiently — including the demonstration FBR (DFBR) programme — a steering committee for R&D was organized in 1986 by the Japan Atomic Power Company, the Power Reactor and Nuclear Fuel Development Corporation, the Japan Atomic Energy Research Institute and the Central Research Institute of Electric Power Industry. A design study of the DFBR is now underway to define its basic specifications by 1990. R&D for Monju, DFBR and future commercial FBRS has been done (1) to improve key technologies developed through the Joyo and Monju programmes; (2) to develop innovative technologies to make FBRS commercial; (3) to promote FBR development in conjunction with the development of the FBR fuel cycle.

Status of Axial Heterogeneous Liquid-Metal Fast Breeder Reactor Core Design Studies and Research and Development

The current status of axial heterogeneous core (AHC) design development in Japan, which consists of an AHC core design in a pool-type demonstration fast breeder reactor (DFBR) and research and development activities supporting AHC core design, is presented. The DFBR core design objectives developed by The Japan Atomic Power Company include (a) favorable core seismic response, (b) core compactness, (c) high availability, and (d) lower fuel cycle cost. The AHC concept was selected as a reference pool-type DFBR core because it met these objectives more suitably than the homogeneous core (HOC). The AHC core layouts were optimized emphasizing the reduction of the burnup reactivity swing, peak fast fluence, and power peaking. The key performance parameters resulting from the AHC, such as flat axial power/flux distribution, lower peak fast fluence, lower burnup reactivity swing, etc., were evaluated in comparison with the HOC. The critical experiments at the Japan Atomic Energy Research Institute’s Fast Critical Assembly facility demonstrate the key AHC performance characteristics. The large AHC engineering benchmark experiments using the zero-power plutonium reactor and the AHC fuel pin irradiation test program using the JOYO reactor are also presented.

SBWR, a simplified boiling water reactor

An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology Supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept.

Development and Operational Experiences of an Automated Remote Inspection System for Interior of Primary Containment Vessel of a BWR

A prototype was developed for an automated remote inspection system featuring continuous monitoring of the working status of major components inside the primary containment vessel of a boiling water reactor. This inspection system consists of four units, or vehicles, which are towed by a trolley chain along a monorail; a complex coaxial cable for data transmission and for power supply; and an operator's console. A TV camera, microphone, thermometer, hygrometer, and ionization chamber are mounted on the various units. After several months' testing under high-ambient temperature, the system was installed in the Tokai-2 power station of Japan Atomic Power Company for in situ tests.