Fabrication and hydrogen generation reaction with water vapor of prototypic pebbles of binary beryllides as advanced neutron multiplier

Nakamichi, Masaru; Kim, Jae-Hwan

Fusion Engineering and Design, 98-99, p.1838 - 1842, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.04.026

Advanced neutron multipliers with high stability at high temperatures are desired for the pebble bed blankets of DEMO reactors. Beryllium intermetallic compounds (beryllides) are the most promising material for this purpose. To fabricate the beryllide pebbles, a new granulation process has been established that combines a plasma sintering method for beryllide synthesis and a rotating electrode method using a plasma-sintered electrode for granulation. In granulation examinations, prototypic pebbles 1 mm in diameter of Be-V beryllide as well as Be-Ti beryllide were successfully fabricated. This study performed not only granulation of binary beryllides but also its characterization of the hydrogen generation reaction with water vapor compared with those of pure Be pebbles.

Analysis of accident scenarios of a water-cooled tokamak DEMO

Nakamura, Makoto; Ibano, Kenzo*; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Ogawa, Yuichi*

Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10

https://conferences.iaea.org/indico/contributionDisplay.py?contribId=397&sessionId=31&confId=46

Of late in Japan, a design study has been undertaken of a tokamak fusion DEMO with pressurized water coolant and solid pebble bed breeding blanket, but safety characteristics of this type of DEMO have not been well examined. In this paper, thermohydraulics analysis of in-vessel and ex-vessel loss-of-coolant accidents of a water-cooled tokamak DEMO is reported. Safety characteristics of water-cooled DEMO, particularly possible loads onto confinement barriers, are discussed based on the thermohydraulics analysis results. Measures to reduce such loads are also proposed.

Elemental development of beryllide electrode for pebble production by rotating electrode method

Uchida, Munenori*; Uda, Minoru*; Iwadachi, Takaharu*; Nakamichi, Masaru; Kawamura, Hiroshi

Journal of Nuclear Materials, 329-333(Part2), p.1342 - 1346, 2004/00

https://doi.org/10.1016/j.jnucmat.2004.04.195

In this paper, an elemental technology to fabricate beryllide rods by the vacuum casting was researched. Furnace material study to prevent the chemical reaction with beryllide and casting procedure study to cast durable ingot without any shrinkages and cracks were performed. From the results of the reactivity test of refractory material with BeTi, it was clear that the BeO crucible had less reactivity with melt and no contamination. From the results of casting tests with a MgO cylindrical mold in a vacuum chamber, it was revealed that the mold dimension was critical to minimize shrinkages and cracks. It was also found that the forced cooling by the MgO cylindrical sleeve with water-cooled copper mold on the bottom was efficient to improve the shrinkages and the cracks.

Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

Yanagi, Yoshihiko*; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto*; Kuroda, Toshimasa*; Kosaku, Yasuo; Ohara, Yoshihiro

Journal of Nuclear Science and Technology, 38(11), p.1014 - 1018, 2001/11

https://doi.org/10.3327/jnst.38.1014

no abstracts in English

Development of HIP bonding procedure and mechanical properties of HIP bonded joints for reduced activation ferritic steel F-82H

; Kurasawa, Toshimasa; Kuroda, Toshimasa*; Hatano, Toshihisa; Takatsu, Hideyuki

JAERI-Tech 97-013, 141 Pages, 1997/03

JAERI-Tech-97-013.pdf:16.98MB

no abstracts in English

Preparation of assessment methodologies of the dose rate due to tritium release to the environment from a fusion DEMO 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.

Progress in DEMO design study and issues

Tobita, Kenji

no journal, , 

Status of DEMO design conducted under the Broader Approach Activity is reported highlighting on Japanese design activity. Remote maintenance is one of the most critical design issues in DEMO in that a reasonable plant availability needs to be attained in severe radiation environment. Recent design study revealed that the application of "sector maintenance" scheme to a medium size DEMO with a major radius of about 8 m would lead to increase in the size of toroidal field coils, the required current of poloidal field coils and the sector weight. For this reason, "banana-type segment" scheme is under study instead of the sector scheme. In order to resolve diverter heat removal problem, use of copper alloy pipes in high heat flux and low dpa (displacements per atom) zones is under study. Besides, a possible management scenario of radioactive waste produced in periodic maintenance and the resulting waste-related facilities for DEMO are presented.

TBM as a test body for the assessment of neutron and tritium, and its problems

Kawamura, Yoshinori

no journal, , 

The most important subject in the ITER-TBM test program is the estimation of TBR based on the information of neutron entered into TBM and of tritium recovered from TBM, and is the showing good prospect of TBR for DEMO. To discuss TBR estimation method in TBM test program, JAEA is carrying out the tritium generation and recovery experiments using simulated blanket at Fusion Neutron Source facility (FNS). At FNS, more accurate TBR estimation is possible by the comparison among the amount of calculated tritium generation, observed tritium generation, and tritium recovery. In the case of TBM, the neutron estimation is not so easy, because neutron source is a heterogeneous plasma having volume, and the plasma is generated in the form of pulses. So, TBR estimation in TBM program will need more effort.

Study of loss-of-coolant events in a water-cooled tokamak DEMO

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Gulden, W.*

no journal, , 

Thermohydraulic responses of the DEMO to in-vessel and ex-vessel loss of the primary coolant have been analyzed by the MELCOR code. The analyses have identified possible event sequences following such accidents, their time scales and loads to the confinement barriers of the radioactive materials. On basis of the analyses results, measures to reduce the loads challenging the confinement barrier are proposed.

Scope of the joint special design team for fusion DEMO

Tobita, Kenji

no journal, , 

no abstracts in English