Performance tests of catalysts for the safe conversion of hydrogen inside the nuclear waste containers in Fukushima Daiichi

Reinecke, E.-A.*; Takenaka, Keisuke*; Ono, Hitomi*; Kita, Tomoaki*; Taniguchi, Masashi*; Nishihata, Yasuo; Hino, Ryutaro; Tanaka, Hirohisa*

International Journal of Hydrogen Energy, 46(23), p.12511 - 12521, 2021/03

https://doi.org/10.1016/j.ijhydene.2020.08.262

The safe decommissioning as well as decontamination of the radioactive waste resulting from the nuclear accident in Fukushima Daiichi represents a huge task for the next decade. At present, research and development on long-term safe storage containers has become an urgent task with international cooperation in Japan. One challenge is the generation of hydrogen and oxygen in significant amounts by means of radiolysis inside the containers, as the nuclear waste contains a large portion of sea water. The generation of radiolysis gases may lead to a significant pressure build-up inside the containers and to the formation of flammable gases with the risk of ignition and the loss of integrity. In the framework of the project "R&D on technology for reducing concentration of flammable gases generated in long-term waste storage containers" funded by the Japanese Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT), the potential application of catalytic recombiner devices inside the storage containers is investigated. In this context, a suitable catalyst based on the so-called intelligent automotive catalyst for use in a recombiner is under consideration. The catalyst is originally developed and mass-produced for automotive exhaust gas purification, and is characterized by having a self-healing function of precious metals (Pd, Pt and Rh) dissolved as a solid solution in the perovskite type oxides. The basic features of this catalyst have been tested in an experimental program. The test series in the REKO-4 facility has revealed the basic characteristics of the catalyst required for designing the recombiner system.

Influence of surface condition on deuterium release from LiTiO pebble

Tsuchiya, Akihito*; Yamauchi, Yuji*; Nobuta, Yuji*; Hino, Tomoaki*; Akiba, Masato; Enoeda, Mikio

Fusion Engineering and Design, 89(7-8), p.1280 - 1283, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.03.054

Lithium titanate (LiTiO) pebbles were irradiated with D3+ ions with energy of 5.0 keV, and the amounts of retained deuterium in the pebbles were measured by thermal desorption spectroscopy. Cycles of irradiation and heating were carried out repeatedly in order to investigate the influence of surface condition on deuterium release from LiTiO. The composition of Li decreased with the increase in the number of the irradiation and heating cycle. The desorption peaks of the gases contained deuterium atoms were shifted to higher temperature region. Total amount of desorbed deuterium showed the tendancy to increase. These results suggest that the tritium recovery efficiency would deteriorate with the increase in the damages and the defects especially at the lower temperature region during the use of blanket.

Effects of hydrogen mixture into helium gas on deuterium removal from lithium titanate

Tsuchiya, Akihito*; Hino, Tomoaki*; Yamauchi, Yuji*; Nobuta, Yuji*; Akiba, Masato; Enoeda, Mikio

Fusion Engineering and Design, 88(9-10), p.2298 - 2301, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.05.028

Lithium titanate (LiTiO) pebbles were irradiated with deuterium ions with energy of 1.7 keV and then exposed to helium or helium and hydrogen mixed gas at various temperatures, in order to evaluate the effects of gas exposure on deuterium removal from the pebbles. The amounts of residual deuterium in the pebbles were measured by thermal desorption spectroscopy. The mixing of hydrogen gas into helium gas enhanced the removal amount of deuterium. In addition, we also evaluated the pebbles exposed to the helium gas with different hydrogen mixture ratio from 0% to 1%, at 573 K. Although the amount of residual deuteriumin the pebbles after the exposure decreased with increasing the hydrogen mixture ratio, the implanted deuterium partly remained after the exposure. These results suggest that the tritium inventory may occur at low temperature region in the blanket during the operation.

Hardness modification of aluminum-alloys by means of energetic ion irradiation and subsequent thermal aging

Mitsuda, Tomoaki*; Kobayashi, Ippei*; Kosugi, Shinya*; Fujita, Naoki*; Saito, Yuichi; Hori, Fuminobu*; Semboshi, Satoshi*; Kaneno, Yasuyuki*; Nishida, Kenji*; Soneda, Naoki*; et al.

Nuclear Instruments and Methods in Physics Research B, 272, p.49 - 52, 2012/02

https://doi.org/10.1016/j.nimb.2011.01.030

Deuterium retention in F82H after low energy hydrogen ion irradiation

Ito, Tatsuya*; Yamauchi, Yuji*; Hino, Tomoaki*; Shibayama, Tamaki*; Nobuta, Yuji*; Ezato, Koichiro; Suzuki, Satoshi; Akiba, Masato

Journal of Nuclear Materials, 417(1-3), p.1147 - 1149, 2011/10

https://doi.org/10.1016/j.jnucmat.2010.12.206

Experimental studies on in-vessel tritium inventory in complicated environment of fusion reactors

Ueda, Yoshio*; Hino, Tomoaki*; Ono, Noriyasu*; Takagi, Ikuji*; Nakano, Tomohide; Tanabe, Tetsuo*; Kajita, Shin*; Fukumoto, Masakatsu

Purazuma, Kaku Yugo Gakkai-Shi, 85(10), p.684 - 694, 2009/10

http://www.jspf.or.jp/Journal/PDF_JSPF/jspf2009_10/jspf2009_10-684.pdf

Relation between deuterium retention, wall temperature and hydrocarbon generation in 30-s H-mode discharges of JT-60U was described. In discharges with a density below 50% of the Greenwald density, the trend of the deuterium retention against pulse number depended on the wall temperature: with increasing pulse number, the deuterium retention decreased at a wall temperature of 300 C, decreased gradually at 150 C, and remained constant at 80 C. In contrast, in discharges with high densities above 70% of the Greenwald density, the deuterium retention increased with increasing pulse number at the above three wall temperatures. In the high density discharges, the deuterium retention flux increased with increasing hydrocarbon generation flux, suggesting that the deuterium is retained in co-deposition layers of carbon, which originated from the hydrocarbons.

Comparison of boronized wall in LHD and JT-60U

Ashikawa, Naoko*; Kizu, Kaname; Yagyu, Junichi; Nakahata, Toshihiko*; Nobuta, Yuji; Nishimura, Kiyohiko*; Yoshikawa, Akira*; Ishimoto, Yuki*; Oya, Yasuhisa*; Okuno, Kenji*; et al.

Journal of Nuclear Materials, 363-365, p.1352 - 1357, 2007/06

https://doi.org/10.1016/j.jnucmat.2007.01.271

no abstracts in English

Characteristics of plasma operation with the ferritic inside wall and its compatibility with high-performance plasmas in JFT-2M

Tsuzuki, Kazuhiro*; Kimura, Haruyuki; Kusama, Yoshinori; Sato, Masayasu; Kawashima, Hisato; Kamiya, Kensaku; Shinohara, Koji; Ogawa, Hiroaki; Uehara, Kazuya; Kurita, Genichi; et al.

Fusion Science and Technology, 49(2), p.197 - 208, 2006/02

https://doi.org/10.13182/FST06-A1095

no abstracts in English

Hydrogen retention in divertor tiles used in JT-60 for hydrogen discharge period

Hirohata, Yuko*; Shibahara, Takahiro*; Tanabe, Tetsuo*; Arai, Takashi; Goto, Yoshitaka*; Oya, Yasuhisa*; Yoshida, Hajime*; Morimoto, Yasutomi*; Yagyu, Junichi; Masaki, Kei; et al.

Journal of Nuclear Materials, 337-339, p.609 - 613, 2005/03

https://doi.org/10.1016/j.jnucmat.2004.10.078

no abstracts in English

Impurity release and deuterium retention properties of a ferritic steel wall in JFT-2M

Ogawa, Hiroaki; Yamauchi, Yuji*; Tsuzuki, Kazuhiro; Kawashima, Hisato; Sato, Masayasu; Shinohara, Koji; Kamiya, Kensaku; Kasai, Satoshi; Kusama, Yoshinori; Yamaguchi, Kaoru*; et al.

Journal of Nuclear Materials, 329-333(Part1), p.678 - 682, 2004/08

https://doi.org/10.1016/j.jnucmat.2004.04.192

no abstracts in English