| 1. |
K. Masuta, Y. Hara, M. Oya, N. Yoshida, K. Katayama, Measurement of Hydrogen Permeation Fluxes Through Tungsten Deposition Layer Growing by Hydrogen Plasma Sputtering and Observation of Microstructure, 2024.01. |
| 2. |
S. Fukada, R. Sakai, M. Oya, K. Katayama, Selective H2 Evolution and CO2 Absorption in Electrolysis of Ethanolamine Aqueous Solutions, 2023.11. |
| 3. |
T. Nagasaka, T.Tanaka, M. Kobayashi, K. Fukumoto, T. Toyama, T. Sugawara, R. Kasada, Y. Yamauchi, K. Katayama, M. Oya, K. Yabuuchi, S. Sakurai, K. Nomura, H. Yoshinaga, Ten-Year Recycling of Vanadium Alloy in Fusion Reactors」Materials Science Forum , 2023.12. |
| 4. |
A. Ipponsugi, K. Katayama, T. Matsumoto, S. Iwata, M. Oya and Y. Someya, Study on Tritium Permeation from the Primary to the Secondary Water Coolant for Fusion Reactors, 2023.10. |
| 5. |
H. Sun, K. Katayama and M. Oya, Contribution of electron density to plasma decomposition rate of methane, 2023.07. |
| 6. |
K. Katayama, K. Kubo, T. Ichikawa, M. Oya, S. Fukada, Y.Iinuma, Tritium release behavior from neutron-irradiated FLiNaK mixed with Ti powder, 2023.05. |
| 7. |
M. Oya, R. Ikeda, K. Katayama, Effective Decomposition of Water Vapor in Radio-Frequency Plasma with Carbon Deposition on Vessel Wall, 10.1585/pfr.17.2405087, 17, 2405087, 2022.07, 核融合炉において、燃料であるトリチウムを効率的に利用するため、排気されるトリチウム化水素化合物ガス（水蒸気やメタンなど）を分解してトリチウムを抽出する必要がある。本研究室では、水素化合物ガスをプラズマで分解する新しい手法を提案した。先行研究は、プラズマによるメタンの分解を実証したが、容器内壁に炭素が堆積することが問題となった。本研究では、水蒸気のプラズマ分解実験を実施するとともに、炭素堆積が分解効率に与える影響を調べた。その結果、炭素堆積により、水蒸気の分解効率が向上することが確認された。. |
| 8. |
Y. Hara, K. Katayama and M. Oya, Modeling of hydrogen permeation behavior through tungsten deposition layer growing on nickel substrate by hydrogen plasma sputtering, Fusion Engineering and Design, 10.1016/j.fusengdes.2021.112851, 172, 112851, 2021.11. |
| 9. |
K. Katayama, Y. Someya, T. Chikada, K. Tobita, H. Nakamura, Y. Hatano, R. Hiwatari, Y. Sakamoto, A. Ipponsugi, M. Oya, Effect of temperature distribution on tritium permeation rate to cooling water in JA DEMO condition, Fusion Engineering and Design, 2021.05. |
| 10. |
K. Hanada, N. Yoshida, M. Hasegawa, M. Oya, Y. Oya, I. Takagi, A. Hatayama, T. Shikama, H. Idei, Y. Nagashima, R. Ikezoe, T. Onchi, K. Kuroda, S. Kawasaki, A. Higashijima, T. Nagata, S. Shimabukuro, K. Nakamura, S. Murakami, Y. Takase, X. Gao, H. Liu, and J. Qian, Overview of recent progress on steady state operation of all-metal plasma facing wall device QUEST, Nuclear Materials and Energy, 2021.05. |
| 11. |
K. Kubo, K. Katayama, M. Oya, K. Tsukahara, S. Fukada, T. Tanaka, A. Sagara, J. Yagi, Y. Iinuma, Tritium release behavior from neutron-irradiated FLiNaBe mixed with titanium powder, Fusion Engineering and Design, 2021.04. |
| 12. |
H. Ito, K. Katayama, D. Mori, Y. Hara and M. Oya, Hydrogen permeation behavior through tungsten deposition layer, Fusion Engineering and Design, 2020.11. |
| 13. |
M. Oya, M. Shimada, C.N. Taylor, M.I. Kobayashi, Y. Nobuta, Y. Yamauchi, Y. Oya, Y. Ueda, Y. Hatano, Deuterium retention in tungsten irradiated by high-dose neutrons at high temperature, Nuclear Materials and Energy, 2021.06. |
| 14. |
Q. Yue, K. Hanada, M. Oya, S. Matsuo, S. Kojima, H. Idei, T. Onchi, K. Kuroda, N. Yoshida, R. Ikezoe, Y. Liu, M. Hasegawa, S. Shimabukuro, A. Higashijima, T. Nagata and S. Kawasaki, Measurement of Dynamic Retention with Fast Ejecting System of Targeted Sample (FESTA), Plasma and Fusion Research, 2020.04. |
| 15. |
M. Oya, Y. Hara, K. Katayama, K. Ohya, Simulation of Experimental Deuterium Retention in Tungsten under Periodic Deuterium Plasma Irradiation, 2021.05. |
| 16. |
M. Oya, R. Ikeda, K. Katayama, Atomic and Molecular Processes in Plasma Decomposition Method of Hydrocarbon Gas, Plasma and Fusion Research, 10.1585/pfr.15.2405032, 15, 2405032, 1-6, 2020.05. |
| 17. |
Makoto Oya, G. Motojima, M. Tokitani, H. Tanaka, S. Masuzaki, Y. Ueda, The role of the graphite divertor tiles in helium retention on the LHD wall, Nuclear Materials and Energy, 10.1016/j.nme.2017.07.006, 13, 58-62, 2017.12, In this study, global particle balance of helium (He) long pulse discharge with graphite divertor in the Large Helical Device (LHD) and He retention in graphite from lab-scale experiments were compared in order to determine the role of the graphite divertor in He retention on the LHD wall. Global He particle balance analysis was conducted in long-pulse discharges in LHD with only turbo molecular pump. The analysis showed that static He retention was ∼2.9 × 1022 He and the ratio of retained He (= retention over fluence) was ∼0.45%. In the lab-scale study, He retention was measured by Thermal Desorption Spectroscopy (TDS) after ion irradiation or plasma exposure. The ratio of retained He were 0.1 ∼ 0.5% in all conditions, which was well consistent with LHD results. Therefore, it may be concluded that graphite divertor tiles have an important role in He absorption in the LHD wall at the initial phase of discharges.. |
| 18. |
Makoto Oya, H. T. Lee, A. Hara, K. Ibano, M. Oyaidzu, T. Hayashi, Y. Ueda, Effect of periodic deuterium ion irradiation on deuterium retention and blistering in Tungsten, Nuclear Materials and Energy, 10.1016/j.nme.2017.03.022, 12, 674-677, 2017.08, The effect of periodic irradiation on Deuterium (D) retention and blistering in Tungsten (W) was investigated. W samples were exposed to D plasma at a fixed fluence while varying the irradiation cycle number (1-shot, 2-shots and 3-shots). Exposure energy and flux were ∼50 eV and ∼1 × 10^22 D m^−2 s^−1, respectively. Sample temperatures were 537 K and 643 K. At 573 K, D retention and blister density decreased with increasing number of irradiation cycle. In contrast at 643 K, D retention showed no dependence on number of irradiation cycle. Therefore, sample temperature during irradiation is an important parameter in comparing the results of continuous and periodic irradiation, especially in studies involving extremely-high-flux (10^24 D m^−2 s^−1) irradiation and fluence dependency of D retention.. |
| 19. |
K. Iwano, K. Yamanoi, Y. Iwasa, K. Mori, Y. Minami, R. Arita, T. Yamanaka, K. Fukuda, M.J.F. Empizo, K. Takano, T. Shimizu, M. Nakajima, M. Yoshimura N. Sarukura, T. Norimatsu, M. Hangyo, H. Azechi, B.G. Singidas, R.V. Sarmago, M. Oya, Y. Ueda, Optical transmittance investigation of 1-keV ion-irradiated sapphire crystals as potential VUV to NIR window materials of fusion reactors, AIP Advances, 2016.10. |
| 20. |
K. Yakushiji, H.T. Lee, M. Oya, Y. Hamaji, K. Ibano, Y. Ueda, Influence of helium on deuterium retention in reduced activation ferritic martensitic steel (F82H) under simultaneous deuterium and helium irradiation, Physica Scripta, 2016.01. |
| 21. |
V.Kh. Alimov, Y. Hatano, K. Sugiyama, B. Tyburska-Pueschel, M. Oya, Y. Ueda, K. Isobe, A. Hasegawa, Influence of He implantation on deuterium trapping at defects induced in W by irradiation with MeV-range W ions, Journal of Plasma Fusion Research, 2015.03. |
| 22. |
Makoto Oya, H. T. Lee, Y. Ueda, H. Kurishita, M. Oyaidzu, T. Hayashi, N. Yoshida, T. W. Morgan, G. De Temmerman, Surface morphology changes and deuterium retention in Toughened, Fine-grained Recrystallized Tungsten under high-flux irradiation conditions, Journal of Nuclear Materials, 10.1016/j.jnucmat.2014.11.124, 463, 1037-1040, 2015.07, Abstract Surface morphology changes and deuterium (D) retention in Toughened, Fine-Grained Recrystallized Tungsten (TFGR W) with TaC dispersoids (W-TaC) and pure tungsten exposed to D plasmas to a fluence of 1026 D/m2 s were studied as a function of the D ion flux (1022-1024 D/m2 s). As the flux increased from 1022 D/m2 s to 1024 D/m2 s, the numbers of blisters increased for both materials. However, smaller blisters were observed on W-TaC compared to pure W. In W-TaC, cracks beneath the surface along grain boundaries were observed, which were comparable to the blister sizes. The reason for the smaller blister sizes may arise from smaller grain sizes of W-TaC. In addition, reduction of the D retention in W-TaC was observed for higher flux exposures. D depth profiles indicate this reduction arises due to decrease in trapping in the bulk.. |
| 23. |
K. Hanada, N. Yoshida, M. Hasegawa, A. Hatayama, K. Okamoto, I. Takagi, T. Hirata, Y. Oya, M. Miyamoto, M. Oya, T. Shikama, A. Kuzmin, Z.X. Wang, H. Long, H. Idei, Y. Nagashima, K. Nakamura, O. Watanabe, T. Onchi, H. Watanabe, K. Tokunaga, A. Higashijima, S. Kawasaki, T. Nagata, S. Shimabukuro, Y. Takase, S. Murakami, X. Gao, H. Liu, J. Qian, R. Raman and M. Ono, Particle balance investigation with the combination of the hydrogen barrier model and rate equations of hydrogen state in long duration discharges on an all-metal plasma facing wall in QUEST, Nuclear Fusion, 2019.05. |
| 24. |
K. Hanada, N. Yoshida, I. Takagi, T. Hirata, A. Hatayama, K. Okamoto, Y. Oya, T. Shikama, Z. Wang, H. Long, C. Huang, M. Oya, H. Idei, Y. Nagashima, T. Onchi, M. Hasegawa, K. Nakamura, H. Zushi, K. Kuroda, S. Kawasaki, A. Higashijima, T. Nagata, S. Shimabukuro, Y. Takase, S. Murakami, X. Gao, H. Liu, J. Qian, R. Raman, and M.Ono, Estimation of fuel particle balance in steady state operation with hydrogen barrier model, Nuclear Materials and Energy, 2019.05. |
| 25. |
M. J. F. Empizo, K. Yamanoi, K. Mori, K. Iwano, Y. Iwasa, Y. Minami, R. Arita, K. Fukuda, K. Takano, T. Shimizu, M. Nakajima, M. Yoshimura, N. Sarukura, T. Norimatsu, M. Hangyo, H. Azechi, T. Fukuda, B. G. Singidas, R. V. Sarmago, M. Oya, Y. Ueda, Optical damage assessment and recovery investigation of hydrogen-ion and deuterium-ion plasma-irradiated bulk ZnO single crystals, 2017.05. |
| 26. |
K. Yakushiji, H.T. Lee, M. Oya, M. Tokitani, A. Sagara, Y. Hamaji, K. Uenishi, K. Ibano, Y. Ueda, Erosion and morphology changes of F82H steel under simultaneous hydrogen and helium irradiation, Fusion Engineering and Design, 2017.03. |
| 27. |
Y. Hamaji, H.T. Lee, A. Kreter, S. Möller, M. Rasinski, M. Tokitani, S. Masuzaki, A. Sagara, M. Oya, K. Ibano, Y. Ueda, R. Sakamoto, Damage and deuterium retention of re-solidified tungsten following vertical displacement event-like heat load, Nuclear Materials and Energy, 2016.11. |
| 28. |
K. Yamanoi, M. J. F. Empizo, K. Mori, K. Iwano, Y. Minami, R. Arita, Y. Iwasa, K. Fukuda, K. Katoh, K. Takano, T. Shimizu, M. Nakajima, N. Sarukura, T. Norimatsu, M. Hangyo, H. Azechi, T. Fukuda, B. G. Singidas, R. V. Sarmago, M. Oya, Y. Ueda, ZnO crystal as the potential damage-recoverable window material for fusion reactors, Optical Materials, 2016.11. |
| 29. |
Y. Ueda, M. Oya, Y. Hamaji, H.T. Lee, H. Kurishita, Y. Torikai, N. Yoshida, A. Kreter, J.W. Coenen, A. Litnovsky, V. Phillips, Surface erosion and modification of toughened, fine-grained, recrystallized tungsten exposed to TEXTOR edge plasma, Physica Scripta, 2014.04. |
| 30. |
Makoto Oya, H. T. Lee, Y. Ohtsuka, Y. Ueda, H. Kurishita, M. Oyaidzu, T. Yamanishi, Deuterium retention in various toughened, fine-grained recrystallized tungsten materials under different irradiation conditions, Physica Scripta, 10.1088/0031-8949/2014/T159/014048, T159, 2014.04, Deuterium retention in two types of toughened, fine-grained recrystallized W (TFGR W-1.2 wt% titanium carbide (TiC) and TFGR W-3.3 wt% tantalum carbide (TaC)) was studied, compared to pure W. D plasma exposure was performed to a fluence of 1 × 1026 D m-2 at a temperature of 573 K, followed by retention measurement analysis by nuclear reaction analysis and thermal desorption spectroscopy (TDS). It is found that D retention in TFGR W is higher than that in pure W. This is because TFGR W has a high density of trapping sites with low trapping energy and dispersoid (TiC or TaC) may serve as additional trapping sites with high trapping energy. Different irradiation experiments (D ion beam implantation) were also conducted at sample temperatures of 473-873 K, followed by TDS. At higher sample temperature (> 700 K), D retention in TFGR W-3.3 wt% TaC is lower than that in TFGR W-1.2 wt% TiC. This may be due to different types of dispersoids.. |
| 31. |
Makoto Oya, K. Uekita, H. T. Lee, Y. Ohtsuka, Y. Ueda, H. Kurishita, A. Kreter, J. W. Coenen, V. Philipps, S. Brezinsek, A. Litnovsky, K. Sugiyama, Y. Torikai, Deuterium retention in Toughened, Fine-Grained Recrystallized Tungsten, Journal of Nuclear Materials, 10.1016/j.jnucmat.2013.01.230, 438, SUPPL, 2013, Deuterium retention in Toughened, Fine-Grained Recrystallized W (TFGR W-1.1 wt%TiC) was studied, compared to pure W. D implantation was performed to a fluence of 1 × 1024m 2 at temperatures of 473-873 K, followed by TDS. It was found that D retention in TFGR W is higher than in pure W at all irradiation temperatures. Namely, at 673 K, D retention in TFGR W is six times higher than pure W. TDS spectrum of TFGR W irradiated at 573 Khas a large peak around ∼700 K with small shoulder up to ∼1100 K. In the case of D + He simultaneous irradiation, D retention is about 30% lower than for pure D. In addition, plasma exposure experiment was also conducted in TEXTOR, followed by NRA. Higher retention in TFGR W-1.1 wt%TiC could be attributed to high grain boundary diffusion (then trapping deeper into the bulk) and formation of TiD2.. |
