Kyushu University Academic Staff Educational and Research Activities Database
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Kazunari Katayama Last modified date:2020.06.30

Associate Professor / Engineering Science for Advanced Energy System
Department of Advanced Energy Science and Engineering
Faculty of Engineering Sciences

Graduate School
Undergraduate School

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 Reseacher Profiling Tool Kyushu University Pure
Academic Degree
doctor of engineering
Field of Specialization
Fusion engineering
Research Interests
  • Study on molten salt circulation and control technology aiming at development of molten salt reactors
    keyword : Fluoride molten salt, tritium, chemical control
  • Fundamental study on tritium behavior in supercritical CO2 gas turbine power generation system
    keyword : Tritium, supercritical CO2
  • A study on tritium production and confinement by high-temperature gas-cooled reactor
    keyword : Tritium, High-temperature gas-cooled reactor
  • Development of tritium transfer model in soil
    keyword : Tritium, soil, percolation, isotope exchange
  • A study on mass transfer and tritiumu behavior in tritium breeding ceramics materials
    keyword : Blanket system in fusion reactor, tritium, Lithium
  • A study on mass transfer phenomena in plasma facing wall of fusion reactors
    keyword : Plasma-wall interaction, tritium, mass transfer
  • Research and development of decomposition-recovery method of tritiated hydrocarbon by plasma
    keyword : High Temperature Gas Cooling Reactor, Tritium processing, Plasma decomposition
  • A study on tritium behavior in liquid blanket materials
    keyword : Tritium, Flibe(2LiF+BeF2)、Lead-Lithium
Academic Activities
1. K. Katayama, N. Ashikawa, F. Ding, H. Mao, H.S. Zhou, G.N. Luo, J. Wu, M. Noguchi, S. Fukada, Deuterium retention in deposited W layer exposed to EAST deuterium plasma, Nuclear Materials and Energy, 12, 617-621, 2017.07,
 The deposited W layers formed on the W plate by hydrogen plasma sputtering were exposed to deuterium plasma in EAST together with bare W plate. In TDS measurement, the deuterium release was clearly observed from the deposited W layer in addition to the release of hydrogen which was incorporated during the sputtering-deposition processes. On the other hand, the release of hydrogen isotope was not detected from the bare W plate. This suggests that the formation of deposited W layers increases tritium inventory in the plasma confinement vessel. Although the thermocouple contacting to the backside of the W plate did not indicate a remarkable temperature rise, deuterium release peaks from the W layer were close to that from the W layer irradiated by 2 keV D2 + at 573 K. It was found by glow discharge optical emission spectrometry analysis that retained deuterium in the W layer has a peak at the depth of 50 nm and gradually decreases toward the W substrate. From X-ray photoelectron spectroscopy analysis, it was evaluated that W oxide existed just at the surface and W atoms in the bulk of deposited W layer were not oxidized. These data suggest that hydrogen isotopes are not retained in W oxide but grain boundaries..
2. Kazunari Katayama, Satoshi Fukada, Direct Decomposition Processing of Tritiated Methane by Helium RF Plasma, Fusion Science and Technology, 71, 3, 426-431, 2017.04,
 With the aim of developing a method for the recovery of tritium from tritium-bearing hydrocarbons, it was shown experimentally that methane can be decomposed directly into hydrogen and carbon in RF plasmas via reactions initiated by electrons. Measurements performed with CH4 and CH3T in a helium RF plasma indicate that the degree of decomposition of CH3T is substantially smaller than that of CH4. This is considered to be caused by a very low concentration of CH3T. It was found that a majority of tritium dissociated from CH3T is retained in the plasma reactor. However, a certain amount of retained tritium could be removed by a discharge-cleaning of oxygen..
3. Kazunari Katayama, Youji Someya, Kenji Tobita, Hirofumi Nakamura, Hisashi Tanigawa, Makoto Nakamura, Nobuyuki Asakura, Kazuo Hoshino, Takumi Chikada, Yuji Hatano, Satoshi Fukada, Estimation of Tritium Permeation Rate to Cooling Water in Fusion DEMO Condition, Fusion Science and Technology, 71, 3, 261-267, 2017.04,
 The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle & Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium two-component system. In the representative recent DEMO condition, the following tritium permeation\rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day..
4. Kazunari Katayama, Hiroki Ushida, Hideaki Matsuura, Satoshi Fukada, Minoru Goto, Shigeaki Nakagawa, Evaluation of Tritium Confinement Performance of Alumina and Zirconium for Tritium Production in a High-Temperature Gas-Cooled Reactor for Fusion Reactors, Fusion Science and Technology, 68, 3, 662-668, 2015.10,
 Tritium production utilizing nuclear reactions by neutron and lithium in a high-temperature gas-cooled reactor is attractive for development of a fusion reactor. From viewpoints of tritium safety and recovery efficiency, tritium confinement is an important issue. It is known that alumina has high resistance for gas permeation. In this study, hydrogen permeation experiments in commercial alumina tubes were conducted and hydrogen permeability, diffusivity and solubility were evaluated. By using obtained data, tritium permeation behavior from an Al2O3-coated Li-compound particle was simulated. Additionally, by using literature data for hydrogen behavior in zirconium, an effect of Zr incorporation into an Al2O3 coating on tritium permeation was discussed. It was indicated that the majority of produced tritium was released through the Al2O3 coating above 500 °C. However, it is expected that total tritium leak is suppressed to below 0.67 % of total tritium produced at 500 °C by incorporating Zr fine particles into the inside of Al2O3 coating, assuming tritium pressure inside particle is kept at the plateau pressure of the Zr hydride generation reaction..
5. Kazunari Katayama, Hideaki Kashimura, Tsuyoshi Hoshino, Masabumi Nishikawa, Hideki Yamasaki, Ishinichiro Ishikawa, Yasuhito Ohnishi, Release behavior of water vapor and mass loss from lithium titanate, Fusion Engineering and Design, 87, 5-6, 927-931, 2012.08.
6. Kazunari Katayama. Satoshi Fukada and Masabumi Nishikawa, Direct decomposition of methane using helium RF plasma, Fusion Engineering and Design, 7-9, 85, 1381-1385, 2010.12.
7. K. Katayama, K.Imaoka, M.Tokitani, M.Miyamoto, M. Nishikawa, S. Fukada, N.Yoshida, Deuterium and helium release and microstructure of tungsten deposition layers formed by RF plasma sputtering, Fusion Science and Technology, 54, 2, 549-552, 2008.08.
8. K. Katayama, K. Imaoka, T. Okamura, M. Nishikawa, Helium and hydrogen trapping in tungsten deposition layers formed by helium plasma sputtering, Fusion Engineering and Design, Volume 82, Issues 15-24, October 2007, Pages 1645-1650, 2007.10.
9. K.Katayama, T. Kawasaki, Y. Manabe, H.Nagase, T. Takeishi, M. Nishikawa, Hydrogen retention in carbon-tungsten co-deposition layer formed by hydrogen RF plasma, Thin Solid Films, 506-507 (2006) 188-191, 2006.05.
10. K. Katayama, T. Takeishi, Y. Manabe, H. Nagase, M. Nishikawa, N. Miya, Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U, Journal of Nuclear materials, 10.1016/j.jnucmat.2004.11.005, 340, 1, 83-92, 340 (2005) 83-92, 2005.04.
11. K. Katayama, M. Nishikawa, T. Takeishi, Isotope exchange reaction between tritiated water and hydrogen on SiC, Journal of Nuclear materials, 10.1016/j.jnucmat.2003.09.002, 323, 1, 138-143, 323 (2003) 138, 2003.11.
Other Educational Activities
  • 2020.02.
  • 2019.12.
  • 2019.11.
  • 2019.09.
  • 2019.02.
  • 2018.12.
  • 2018.11.
  • 2018.09.
  • 2018.02.
  • 2017.12.
  • 2017.12.
  • 2017.10.
  • 2017.09.
  • 2017.07.
  • 2017.02.
  • 2016.11.