T Otsuka¹, S Masuzaki², N Ashikawa^2,3, Y Torikai⁴, Y Hatano⁵, M Tokitani^2,3, Y Oya⁶, N Asakura⁷, T Hayashi⁷, H Tanigawa⁷, Y Iwai⁷, A Widdowson⁸, M Rubel⁹ and JET Contributors<sup>10

1</sup> Department of Electric and Electronic Engineering, Kindai University, Higashi-Osaka, Japan
² National Institute for Fusion Science, Toki, Japan
³ The Graduate University for Advanced Studies, SOKENDAI, Toki, Japan
⁴ Graduate School of Science and Engineering, Ibaraki University, Mito, Japan
⁵ Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan
⁶ College of Science, Shizuoka University, Shizuoka, Japan
⁷ National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Japan
⁸ Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, United Kingdom
⁹ KTH Royal Institute of Technology, Stockholm, Sweden
Author note
¹⁰ See author list: J. Mailloux et al 28th IAEA Fusion Energy Conference, 2021.

Abstract
Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015–2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011–2012 (ILW-1). The dominant component in ILW-1 dust was carbon (C) originating from tungsten-coated carbon fibre composite (CFC) tiles in JET-ILW divertor and/or legacy of C dust after the JET operation with carbon wall. Around 85% of the total tritium retention in ILW-1 dust was attributed to the C dust. The retention in tungsten (W) and beryllium (Be) dominated particles was 100 times smaller than the highest T retention in carbon-based particles. After ILW-3 the main component contributing to the T retention was W. The number of small W particles with T increased, in comparison to ILW-1, most probably by the exfoliation and fragmentation of W coatings on CFC tiles though T retention in individual W particles was smaller than in C particles. The detection of only very few Be-dominated dust particles found after ILW-1 and ILW-3 could imply stable Be deposits on the divertor tiles.

Accepted:17 December 2021

Jing Wang^a, Yuji Hatano^a, Takeshi Toyama^b, Tomoaki Suzudo^b,c, Tatsuya Hinoki^d, Vladimir Kh.Alimov^e, Thomas Schwarz-Selinger<sup>f

a</sup> Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan
^b Institute for Materials Research, Tohoku University, Oarai 311-1313, Japan
^c Center for Computational Science and e-Systems, Japan Atomic Energy Agency, Tokai Mura 319-1195, Japan
^d Open Innovation Institute, Kyoto University, Uji 611-0011, Japan
^e Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russia
^f Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, Garching D-85748, Germany

Abstract
To study the effect of the content of chromium (Cr) in the tungsten (W) matrix on the vacancy formation and retention of hydrogen isotopes, the samples of the W-0.3 at.% Cr alloy were irradiated with 6.4 MeV Fe ions in the temperature range of 523–1273 K to a damage level of 0.26 displacement per atom (dpa). These displacement-damaged samples were exposed to D2 gas at a temperature of 673 K and a pressure of 100 kPa to decorate ion-induced defects with deuterium. The addition of 0.3 at.% Cr into the W matrix resulted in a significant decrease in the retention of deuterium compared to pure W after irradiation especially at high temperature (≥773 K). Positron lifetime in W-0.3 at.% Cr alloy irradiated at 1073 K was almost similar to that for non-irradiated one. These facts indicate the suppression of the formation of vacancy-type defects (monovacancies and vacancy clusters) by 0.3 at.% Cr addition, which leads to the significant reduction in deuterium retention in W-0.3 at.% Cr alloy.

Keywords:Tungsten, Chromium, Deuterium retention, Ion irradiation
Accepted:30 November 2021

Keisuke Iwano¹, Jiaqi Zhang¹, Akifumi Iwamoto^1,2, Yuki Iwasa³, Keisuke Shigemori¹, Masanori Hara⁴, Yuji Hatano⁴, Takayoshi Norimatsu¹,Kohei Yamanoi<sup>1

1</sup> Institute of Laser Engineering, Osaka University, 2‑6 Yamadaoka, Suita, Osaka 565‑0871, Japan.
² National Institute for Fusion Science, 322‑6 Oroshi, Toki, Gifu 509‑5292, Japan.
³ National Metrology Institute of Japan(NMIJ), National Institute of Advanced Industrial Science and Technology (AIST), 1‑1‑1 Central 3, Umezono,Tsukuba, Ibaraki 305‑8563, Japan.
⁴ Faculty of Science, Academic Assembly, University of Toyama, 3190 Gofuku, Toyama 930‑8555, Japan.

Abstract
Physical properties of tritium (T) and deuterium (D) have been of great interest as a fuel for nuclear fusion. However, several kinds of the physical properties in a cryogenic environment have not been reported. Optical properties in liquid and solid phases are indispensable for the quality control of the DT fuel. We study the dependence of the refractive index of solid DT on temperature. A dedicated cryogenic system has been developed and forms a transparent solid DT in a prism cell. Refractive index measurements based on Snell’s law were conducted. The refractive indexes of solid DT are from 1.1618 ± 0.0002 to 1.1628 ± 0.0002 in the temperature range of 19.40 K to 17.89 K.

Accepted:11 January 2022

T. Muroga^a, Y. Hatano^b, D. Clark^c, Y. Katoh<sup>d

a</sup> National Institute for Fusion Science, Toki, 509-5292 Japan
^b University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^c U.S. Department of Energy, Germantown, 20874, MD, USA
^d Oak Ridge National Laboratory, Oak Ridge, 37831, TN, USA

Abstract
The Joint Projects under the Japan-USA Fusion Cooperation Program started in 1981 and has continued for more than 40 years. In the Joint Projects, although a wide range of fusion materials and engineering issues were covered, neutron radiation effects on fusion reactor materials have always been the major research emphases, and the neutron irradiation facilities in the US were jointly used by Japanese and US researchers. Japanese test facilities including neutron and charged particle irradiation facilities were complementarily used. The initial focus of the Joint Projects was on fundamental fusion neutron radiation effects and irradiation correlation. Systematic comparison of fission and fusion radiation effects in comparable damage levels and the effects of transmutation-induced helium were investigated. The collaboration was then focused on the effect of dynamic irradiation effects in variable conditions. In addition to the relatively fundamental studies, the Joint Projects contributed largely to development of candidate materials such as RAFM steels, vanadium alloys, SiC/SiC composites, and tungsten alloys, through a mechanism-oriented approach. The Joint Projects also covered issues specific to materials application to fusion blankets and plasma-facing components, including neutron radiation effects such as tritium retention and permeation of neutron-irradiated plasma-facing materials. Various irradiation technologies were developed and applied to the irradiation experiments, including those for in-situ testing. Considering that high energy neutron sources, such as A-FNS and IFMIF-DONES, now have high viability, the research supporting the neutron source programs is essential. The knowledge obtained through the Joint Projects is valuable and should be advanced for this purpose. To this end, it is of urgent necessity to launch an international scientific program accumulating knowledge of fusion neutron radiation effects, including their fundamental aspects.

Keywords: Neutron radiation effects, Irradiation correlation, Low activation materials, Transmutation
Accepted: 20 December 2021

Satoshi Akamaru^a, Haruya Yamamoto^b, Masanori Hara<sup>a

a</sup> Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^b Graduate School of Science and Engineering for Education, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract
The anomalous Hall effect of thin PdCo films (Pd_0.8Co_0.2 alloy films; 5, 15, and 30 nm thick) subjected to various hydrogen concentrations in the gas phase was investigated. The Hall voltage of the 15 and 30 nm thick PdCo films against an external magnetic field exhibited hysteresis, indicating the anomalous Hall effect of PdCo. The hydrogen absorption in the 5 nm PdCo film decreased the Hall voltage in all considered magnetic fields. Moreover, the slope of the Hall voltage around a zero magnetic field decreased. When N₂ was used as the carrier gas, the slope was proportional to the logarithm of the hydrogen concentration. For dry air, the slope differed from and was similar to that for N₂ below and above hydrogen concentrations of 0.5% and 1.0%, respectively. The adsorbed oxygen on the PdCo surface disturbed the dissolved hydrogen in PdCo at low hydrogen concentrations in dry air.

Keywords: PdCo thin film, Magnetism, Hydrogen sensor, Anomalous Hall effect, Adsorbed oxygen
Accepted: 8 December 2021

Yuji Hatano¹, Hiroto Shimoyachi², Tatsuya Asano², Takahiro Kenmotsu³, Takuro Wada⁴, Yasuhisa Oya⁴, Hiroaki Nakamura⁵, Susumu Fujiwara<sup>6

1</sup>Hydrogen Isotope Research Center, University of Toyama
²Graduate School of Science and Engineering, University of Toyama
³Faculty of Life and Medical Sciences, Doshisha University
⁴Graduate School of Integrated Science and Technology, Shizuoka University
⁵National Institute for Fusion Science
⁶Faculty of Materials Science and Engineering, Kyoto Institute of Technology

Abstract
The goal of this study is to establish a simple experimental system to examine the rate of double strand breaks (DSBs) of genome-sized DNA molecules under irradiation of β-rays from tritium under well-controlled conditions for the validation of computer simulation on interactions of biomolecules and ionizing radiation. Irradiation effects were insignificant at tritium concentration of 1300 Bq/cm³, indicating that the effects of β-rays were far smaller than those of oxidation and/or thermal motion at the low dose rate (4.3 μGy/h). Clear increase in DSB rate was observed at tritium concentrations of 3.0–4.0 MBq/cm³. The temperature dependence of DSB rate was examined by using the high concentration tritiated water.

Keywords: Tritium, DNA, Double strand break, Single molecule observation method
Accepted: 29 May 2022

Xiao‑Ou Yi¹, Tatsuya Kuwabara², Vladimir Kh. Alimov³, Yu‑Feng Du⁴, Wen‑Tuo Han¹, Ping‑Ping Liu¹, Bin‑You Yan⁵, Jiu‑Peng Song⁶, Kenta Yoshida⁴, Takeshi Toyama⁴, Fa‑Rong Wan¹, Somei Ohnuki¹, Yuji Hatano⁷, Yasuyoshi Nagai<sup>4

1</sup>School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083, China
²Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8063, Japan
³Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russia
⁴Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313, Japan
⁵Xiamen Tungsten Co. Ltd., Xiamen 361021, China
⁶School of Materials Science and Engineering, Xihua University, Chengdu 610039, China
⁷Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract
Samples of ultra-high-purity tungsten prepared using chemical vapour deposition (CVD) technique were irradiated with neutrons at temperatures T_irr=373–483 K (stage II of defect recovery) and T_irr=573–673 K (stage III) up to 0.15 displacements per atom (dpa) in the Belgian reactor (BR2). The study of the microstructure of neutron-damaged samples using transmission electron microscopy (TEM) revealed visible defects with a predominance of dislocation loops. With an increase in the neutron irradiation temperature, the spatial distribution of the loops acquired pronounced inhomogeneity, and their average size moderately increased. Cavities and voids were not observed. Irradiation-induced hardening was found and a linear correlation was obtained between Vickers microhardness and nanohardness for undamaged and neutron-irradiated CVD-W samples. Irradiation of tungsten with neutrons led to a signifcant increase in the retention of deuterium, which accumulated mainly in vacancy-type traps. Furthermore, the infuence of the columnar grain structure in low-dose neutron-irradiated tungsten seemed to be non-trivial upon deuterium retention.

Keywords: CVD-W · Neutron irradiation · Microstructure · Hardening · Deuterium retention
Accepted: 20 June 2022

Y.Nobuta^a, T.Toyama^b, A.Matsumoto^c, M.Shimada^d, Y.Oya^e, K.Inoue^b, Y.Nagai^b, Y.Hatano<sup>f

a</sup>Graduate School of Engineering, Hokkaido University, Sapporo, 060-0808, Japan
^bInstitute for Materials Research, Tohoku University, Oarai, 311-1313, Japan
^cGraduate School of Science and Engineering, University of Toyama, Toyama, 930-8555, Japan
^dFusion Safety Program, Idaho National Laboratory, Idaho Falls, 83415, USA
^eRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, 422-8529, Japan
^fHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama, 930-8555, Japan

Abstract
The effects of rhenium (Re) addition on deuterium (D) retention in neutron-irradiated tungsten (W) were investigated. Pure W and W-5Re (5 at.%) alloy samples were irradiated with neutrons at High Flux Isotope Reactor using MFE-RB-19 J capsule. The sample temperature and the damage level were 864 K and 0.35 dpa for pure W and 792 K and 0.26 dpa for W-5Re alloy. A portion of the samples was exposed to D plasma at Tritium Plasma Experiment at Idaho National Laboratory at 823 K to a fluence of 5 × 10²⁵ m⁻². Vacancy-type defects in neutron-irradiated samples were examined using positron annihilation spectroscopy (PAS); D retention after plasma exposure was evaluated by thermal desorption spectrometry (TDS).
TDS measurements revealed that D retention in the neutron-irradiated W-5Re alloy was similar to that in the unirradiated W sample, whereas a significant increase in D retention was observed in neutron-irradiated W. Thus, Re addition significantly suppressed the increase in D retention after neutron irradiation. This effect was attributed to the suppression of vacancy-type defect formation, as confirmed by PAS.

Keywords: Plasma facing materials, Tungsten, Tungsten-rhenium, Hydrogen retention, Thermal desorption spectrometry, Positron annihilation spectroscop
Accepted: 1 May 2022

Nesrin O. Cetiner^a, Yuji Hatano^b, Joel L. McDuffee^a, Dan Ilas^a, Yutai Katoh^a, Josina W. Geringer^a, Takeshi Toyama<sup>c

a</sup>Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA
^bUniversity of Toyama, Toyama 930-8555, Japan
^cTohoku University, Oarai 311-1313, Japan

Abstract
Neutron irradiation of tungsten with and without the presence of hydrogen is needed to understand the influence of hydrogen on microstructure development under fusion reactor conditions. However, there is a risk of ignition if air ingress occurs during seal welding of irradiation capsules in a ressurized hydrogen environment. Therefore, an irradiation capsule was designed that contains several disks of vanadium hydrides at a 30% hydrogen-to-metal atomic ratio. During irradiation, the hydrogen is released from the hydrides as the capsule temperature increases, so the irradiation capsule environment is mostly hydrogen when the capsule reaches its 400 ^◦C design temperature. This paper describes the design and operating characteristics of this first-of-a-kind irradiation capsule.

Keywords: Hydrogen retention, Tungsten irradiation, HFIR, HFIR irradiation experiment, Hydrogen infused vanadium, Fusion reactors
Accepted: 2 March 2022