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Osawa, Kazuhito*; Toyama, Takeshi*; Hatano, Yuji*; Yamaguchi, Masatake; Watanabe, Hideo*
Journal of Nuclear Materials, 527, p.151825_1 - 151825_7, 2019/12
Times Cited Count：8 Percentile：66.68(Materials Science, Multidisciplinary)no abstracts in English
Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*
Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06
Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.
Ishizawa, Akihiro*; Idomura, Yasuhiro; Imadera, Kenji*; Kasuya, Naohiro*; Kanno, Ryutaro*; Satake, Shinsuke*; Tatsuno, Tomoya*; Nakata, Motoki*; Nunami, Masanori*; Maeyama, Shinya*; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 92(3), p.157 - 210, 2016/03
The high-performance computer system Helios which is located at The Computational Simulation Centre (CSC) in The International Fusion Energy Research Centre (IFERC) started its operation in January 2012 under the Broader Approach (BA) agreement between Japan and the EU. The Helios system has been used for magnetised fusion related simulation studies in the EU and Japan and has kept high average usage rate. As a result, the Helios system has contributed to many research products in a wide range of research areas from core plasma physics to reactor material and reactor engineering. This project review gives a short catalogue of domestic simulation research projects. First, we outline the IFERC-CSC project. After that, shown are objectives of the research projects, numerical schemes used in simulation codes, obtained results and necessary computations in future.
Knaster, J.*; Ibarra, A.*; Ida, Mizuho*; Kondo, Keitaro; Kikuchi, Takayuki; Ohira, Shigeru; Sugimoto, Masayoshi; Wakai, Eiichi; Watanabe, Kazuhito; 58 of others*
Nuclear Fusion, 55(8), p.086003_1 - 086003_30, 2015/08
Times Cited Count：60 Percentile：95.24(Physics, Fluids & Plasmas)The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, has accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept. In the Test Cell design, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction. The released IFMIF Intermediate Engineering Design Report, which could be complemented if required concurrently with the outcome of the on-going EVA carried out since the entry into force of IFMIF/EVEDA in June 2007, will allow the decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community.
Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10
Times Cited Count：13 Percentile：70.2(Nuclear Science & Technology)After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Watanabe, Tomoo; Ozawa, Kazumasa; Otsuka, Jiro; SASAKI, Kazuhito; Sawada, Makoto
JNC TN4410 2004-003, 20 Pages, 2004/07
JNC-TN4410-2004-003.pdf:0.66MB
The FBR cycle training facility consists of sodium handling training facility and maintenance training facility, and is being contributed to train for the operators and maintenance workers of the prototype fast breeder reactor "Monju". So far, some training courses have been added to the both training courses of sodium handling technologies maintenance technologies in every year in order to carry out be significant training for preparation of Monju restarting. As encouragement of the sodium handling technology training in 2003FY, the sodium heat transfer basic course was equipped as the 9th sodium handling training course with the aims of learning basic principal technology regarding sodium heat transfer. While, for the maintenance training course, a named "Monju Systems Learning Training Course", which aims to learn necessary knowledge as the engineers related Monju development, was provided newly in this year as an improvement concerned the maintenance course. In 2003FY, nine sodium handling technology training courses were carried out total 33 times and 235 trainees took part in those training courses. Also, nine training courses concerning the maintenance technology held 15 times and total 113 trainees participated. On the other hand, the 4th special lecture related sodium technology by France sodium school instructor was held on Mar. 15-17 and 34 trainees participated. Consequently, a cumulative trainees since October in 2000 opened the FBR cycle training facility reached to 1,236 so for.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Araki, Takao*; Watanabe, Kazuhito*; Kittaka, Daigo*; Ishii, Kyoko*; Matsumiya, Hisato*
no journal, ,
Recent findings on safety characteristics of a tokamak DEMO reactor are reported in the case where all the coolant water is lost completely and instantaneously. Assuming that there are neither off-site power nor active emergency cooling, we have analyzed temporal histories of the temperatures of the reactor components using the fusion reactor thermo-hydraulic analysis code MELCOR-fus. We have found that even in such an extremely severe case, the temperatures of the vacuum vessel and in-vessel components do not reach their melting points.
Kudo, Hironobu; Watanabe, Kazuhito; Hiwatari, Ryoji; Asakura, Nobuyuki; Tokunaga, Shinsuke; Someya, Yoji; Nozawa, Takashi; Tanigawa, Hiroyasu
no journal, ,
In a fusion demo reactor, a rump-up scenario of the plasma is studied. The plasma is growing up and contacting on the first wall surface (limiter-phase) before shifting to diverter phase. Heat load of this time is a transient thing of the dozens seconds. However, it is bigger than the heat load which the first wall receives at the steady state. Therefore there are two idea to be taken with a demo reactor for this heat load. One is that addition a function of limiter to blanket oneself. Another one is design limiter as the independent structure. It is necessary to finally compare the superiority and inferiority of both in TBR (Tritium Breeding ratio) influenced by thickness of the surface tungsten layer and occupation area. This study perform conceptual design of independent limiter.
Masui, Akihiro; Nakamura, Makoto; Watanabe, Kazuhito; Someya, Yoji; Tanigawa, Hisashi; Tobita, Kenji
no journal, ,
When a pipe of tokamak cooling system breaks outside of vacuum vessel, it is necessary to keep integrity of buildings to prevent that tritium and ACP contained in coolant are released to environment. A load to buildings is affected by plant design parameters, for example mass of coolant, pipe diameter and volume of buildings. The impacts of these parameters on the load to buildings are shown based on thermal hydraulic analyses.
Hiwatari, Ryoji; Watanabe, Kazuhito; Aoki, Akira; Tobita, Kenji; Demo Design Joint Special Team
no journal, ,
The conceptual design activity for a fusion Demo plant by the DEMO design joint special team has been started according to the report by the Joint-Core Team for the establishment of technology based required for the development of a Fusion DEMO reactor. One of the main subjects will be the operation plan for DEMO in the intermediate check and review around 2020. In this presentation, the present result on the operation plan for DEMO is reported. Three categories are considered; (1) Operation plan for demonstration of electric generation, (2) Operation plan for demonstration of feasibility of fusion energy, (3) Data acquisition plan. From those view points, experimental subjects, technical skill and acquisition data are listed up as for core plasma, in-vessel components (blanket and divertor), fuel cycle system, plant operation, remote maintenance and inspection, safety system, environmental effect. A preliminary operation plan for DEMO based on the Monju operation plan will be reported.
Watanabe, Kazuhito; Nakamura, Makoto; Someya, Yoji; Masui, Akihiro; Katayama, Kazunari*; Hayashi, Takumi; Yanagihara, Satoshi*; Konishi, Satoshi*; Yokomine, Takehiko*; Torikai, Yuji*; et al.
no journal, ,
In the DEMO design, the blanket primary cooling system involves high temperature pressurized water (~300
C). This means the temperature of blanket structural material is higher than that of ITER. This increases tritium permeation ratio from the fusion plasma and blanket breeder to the primary cooling water. Therefore, we need to consider installation of a water detritiation system. In this study, we estimate the demand of water detritiation system from the view point of the amount of tritium permeated to primary cooling water that assumed conservatively. We also organize the issues for management of tritiated water from the other point of view based on the characteristic of the fusion DEMO reactor. The result shows that the existing facilities can be adopted to the DEMO if we can control the tritium ratio of primary cooling water as same as that of CANDU reactor.
Nakamura, Makoto; Tobita, Kenji; Tanigawa, Hisashi; Someya, Yoji; Masui, Akihiro; Watanabe, Kazuhito; Konishi, Satoshi*; Torikai, Yuji*
no journal, ,
Tritium is major radioactive material in a fusion reactor. Evaluation of the dose due to the tritium is essential to understand environmental consequences of incidental or accidental conditions postulated in the fusion reactor. A purpose of this study is to identify issues to apply UFOTRI, a code of tritium dose analysis being used for the ITER safety assessment, the Japanese environmental conditions. Extensive scans of UFOTRI calculation runs were performed in various meteorological and release conditions. The scans show that the contribution of the secondary tritium release is more significant in the cases of lower release height, lesser stable atmosphere or more distant conditions. The analysis, thus, suggests that it is important to take into account the contribution of the secondarily released tritium in evaluating the early dose to the public due to the tritium release.
Tanigawa, Hisashi; Nakamura, Makoto; Someya, Yoji; Masui, Akihiro; Watanabe, Kazuhito
no journal, ,
Blanket in fusion nuclear reactor has three functions such as neutron shielding, heat recovery and tritium breeding. The blanket in ITER is called as Shielding blanket because it has only a function of neutron shielding. The blanket with heat recovery and tritium breeding functions will be equipped with fusion DEMO reactor for the first time. Therefore it is important to study event sequences and their effects related to the blanket for understandings of safety characteristics in DEMO plants. For both of ITER and DEMO, it is the same that a vacuum vessel is treated as the first container against radioactive materials. Safety-related events including in-vessel LOCA and in-box LOCA are analyzed and their effects on the vacuum vessel are assessed.
