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Journal Articles

Simple 3D PIC analysis for beam phase space oscillation in RF driven negative hydrogen ion source

Shibata, Takanori*; Shinto, Katsuhiro; Nakano, Haruhisa*; Hoshino, Kazuo*; Miyamoto, Kenji*; Okoshi, Kiyonori; Nammo, Kesao*; Ikegami, Kiyoshi*; Kawai, Isao*; Oguri, Hidetomo; et al.

Journal of Physics; Conference Series, 2743, p.012007_1 - 012007_5, 2024/05

https://doi.org/10.1088/1742-6596/2743/1/012007

Times Cited Count：0 Percentile：0.00(Physics, Particles & Fields)

Oscillation of the negative hydrogen ion (H) beam phase space in Radio Frequency (RF) ion source is investigated by a simple 3D Particle-In-Cell (PIC) model which takes into account the transport processes of electron, proton and H in the extraction region. The calculation domain is in vicinity of the single beam aperture in J-PARC ion source configuration. In order to understand relation between the plasma density oscillation and the extracted H beam characteristics, the input electron and proton fluxes from the driver region are varied parametrically with the 1st and the 2nd harmonics of the J-PARC RF frequency (2 or 4 MHz). The numerical results give an idea to the main physical processes between the oscillations of the plasma parameters and the extracted H ion trajectories in the different RF phases. Countermeasures to reduce the oscillation mechanisms are also discussed in the presentation.

Journal Articles

First-principles calculation of mechanical properties of simulated debris ZrU $_{1-x}$ O

Itakura, Mitsuhiro; Nakamura, Hiroki; Kitagaki, Toru; Hoshino, Takanori; Machida, Masahiko

Journal of Nuclear Science and Technology, 56(9-10), p.915 - 921, 2019/09

https://doi.org/10.1080/00223131.2019.1604271

Times Cited Count：2 Percentile：18.85(Nuclear Science & Technology)

To elucidate the mechanical properties of fuel debris inside the Fukushima Daiichi Nuclear Power Plant, we use first-principles calculations to evaluate mechanical properties of cubic Zr $_{x}$ U $_{1-x}$ O $_{2}$ , which is a main component of the fuel debris. We focus on the dependence of mechanical properties on the fraction x of zirconium, compare our results with recent experiment of simulated debris, in which dependences of elastic moduli and fracture toughness on the ZrO $_{2}$ content showed deviation from a simple linear relation. We show that elastic moduli drop at around x=0.25 and increase again for larger values of x, as has been observed in experiments. The reason of the drop is a softening owing to disordered atomistic structures induced by the solute zirconium atoms. We also find that stress-strain curves for the x=0.125 case show marked hysteresis owing to the existence of many meta-stable states. We show that this hysteresis leads to slightly increased fracture toughness, but it is not enough to account for the significant increase of fracture toughness observed in experiments.

Journal Articles

Mechanical properties of cubic (U,Zr)O $_{2}$

Kitagaki, Toru; Hoshino, Takanori; Yano, Kimihiko; Okamura, Nobuo; Ohara, Hiroshi*; Fukasawa, Tetsuo*; Koizumi, Kenji

Journal of Nuclear Engineering and Radiation Science, 4(3), p.031011_1 - 031011_7, 2018/07

https://doi.org/10.1115/1.4039847

Journal Articles

Mechanical properties of fuel debris for defueling toward decommissioning

Hoshino, Takanori; Kitagaki, Toru; Yano, Kimihiko; Okamura, Nobuo; Ohara, Hiroshi*; Fukasawa, Tetsuo*; Koizumi, Kenji

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

Journal Articles

R&D status on water cooled ceramic breeder blanket technology

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.

Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.01.035

Times Cited Count：21 Percentile：82.43(Nuclear Science & Technology)

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.

Journal Articles

Fission product separation from seawater by electrocoagulation method

Kitagaki, Toru; Hoshino, Takanori; Sambommatsu, Yuji; Yano, Kimihiko; Takeuchi, Masayuki; Igarashi, Takeshi*; Suzuki, Tatsuya*

Journal of Radioanalytical and Nuclear Chemistry, 296(2), p.975 - 979, 2013/05

https://doi.org/10.1007/s10967-012-2120-4

Times Cited Count：4 Percentile：31.25(Chemistry, Analytical)

Journal Articles

Development of the water cooled ceramic breeder test blanket module in Japan

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; et al.

Fusion Engineering and Design, 87(7-8), p.1363 - 1369, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.03.007

Times Cited Count：36 Percentile：91.71(Nuclear Science & Technology)

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. Fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

JAEA Reports

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m $^{2}$ . This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Journal Articles

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

https://doi.org/10.1088/0029-5515/49/7/075029

Times Cited Count：140 Percentile：97.65(Physics, Fluids & Plasmas)

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m $^{2}$ or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).

Journal Articles

R&Ds of a LiTiO pebble bed for a test blanket module in JAEA

Tanigawa, Hisashi; Hoshino, Tsuyoshi; Kawamura, Yoshinori; Nakamichi, Masaru; Ochiai, Kentaro; Akiba, Masato; Ando, Masami; Enoeda, Mikio; Ezato, Koichiro; Hayashi, Kimio; et al.

Nuclear Fusion, 49(5), p.055021_1 - 055021_6, 2009/05

https://doi.org/10.1088/0029-5515/49/5/055021

Times Cited Count：23 Percentile：63.54(Physics, Fluids & Plasmas)

This paper presents recent achievements of the research activities for the TBM being developed in JAEA, focusing on the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of LiTiO has been improved by LiO additives. In order to analyze the pebble bed behaviour, thermo-mechanical properties of the LiTiO pebble bed has been experimentally obtained. In order to verify nuclear properties of the pebble bed, the activation foil method has been proposed and a preliminary experiment has been conducted. For the tritium behaviour, the chemical densified coating method has been well developed and tritium recovery system has been modified taking account of the design change of the TBM.

Patent

液中からの放射性元素の分離・除去方法

北垣徹; 星野貴紀; 三本松勇二; 矢野公彦; 竹内正行

not registered

JP, 2012-002095 Patent licensing information Patent publication (In Japanese)

【課題】高効率で迅速な凝集分離性能を有し、ポリマー等の有機化合物を使用せず、かつ、添加剤を極力低減することのできる実用化に適した液中からの放射性元素の分離・除去方法を提供する。【解決手段】放射性元素を含む液にその放射性元素を吸着する吸着剤を添加して所定時間攪拌する吸着剤添加・攪拌工程と、その吸着剤添加・攪拌工程後に当該液を電解する電解工程と、その電解工程後に当該液中の固形分を分離する固形分分離工程を含むことを特徴とする。

Oral presentation

Recent status of WCCB TBM development in Japan

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; et al.

no journal, ,

ITER Test Blanket Module (TBM) program is the first module scale breeding blanket performance test in the real DT fusion reaction environment which will be realized in ITER. Therefore, it is regarded as one of the most important development step of breeding blanket for DEMO. For the TBM program, Japan has the plan to test Water Cooled Ceramic Breeder (WCCB) TBM as the primary candidate breeding blanket. Keeping consistency with ITER construction activities, TBM design and development activities are being performed. This paper overviews the recent achievements of the development of the module fabrication technology and tritium production technology of the WCCB TBM in Japan.

Oral presentation

FaCT phase-I evaluation on the advanced aqueous reprocessing process, 3; Highly effective dissolution for FBR MOX fuels

Ikeuchi, Hirotomo; Shibata, Atsuhiro; Sano, Yuichi; Hoshino, Takanori; Suganuma, Takashi; Washiya, Tadahiro

no journal, ,

no abstracts in English

Oral presentation

Study on application of aqueous treatment for fuel debris in post severe accident, 1; Fundamental test on dissolution of uranium and zirconium oxide solid solution by nitric acid

Hoshino, Takanori; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro; Fukasawa, Tetsuo*; Ohara, Hiroshi*

no journal, ,

no abstracts in English

Oral presentation

Feasibility study on application of aqueous treatment for fuel debris in post severe accident, 2; Suggestion on solubilization of fuel debris for nitric acid by solution treatment with UO $_{2}$

Yano, Kimihiko; Hoshino, Takanori; Kaji, Naoya; Washiya, Tadahiro; Fukasawa, Tetsuo*; Ohara, Hiroshi*

no journal, ,

A part of study on the management of fuel debris in unit 1-3 at Fukushima Dai-ichi Nuclear Power Station, it is suggested that solubilization of fuel debris for nitric acid by solution treatment with UO $_{2}$ . It was experimentally confirmed that this process made insoluble (U,Zr)O $_{2}$ , one of the major components, be dissoluble.

Oral presentation