V.Kh. Alimov ^a,b, Y. Hatano^a, N. Yoshida^c, H. Watanabe^c, M. Oyaidzu^d, M. Tokitani^e, T. Hayashi<sup>d

a</sup>Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^bNational Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow 115409, Russia
^cResearch Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580, Japan
^dFusion Research and Development Directorate, International Fusion Energy Research Center, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japan
^eNational Research Institute for Fusion Science, Toki 509-5292, Japan

Abstract

Targets prepared from Reduced Activation Ferritic Martensitic (RAFM) steel F82H were exposed to low- energy (200 eV) deuterium (D) plasma at various temperatures with highfluxes of about 10²² D/m ² s to various fluences in the range from 10²⁵ to 2.5 ×10²⁶ D/m² . Under the plasma exposure, micro-structured layers are formed on the target surfaces, and the surface morphology is dependent on the exposure tem- perature. The erosion yield of the F82H samples increases by a factor of about two as the exposure tem- perature rises in the range from 403 to 773 K.

Keywords: Deuterium plasma, Reduced activation ferritic martensitic steel, Surface morphology, Sputtering erosion

Accepted: 28 January 2016

Masao Matsuyama, Shinsuke Abe

Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

Retention and release behavior of tritium charged into stainless steel type 316 by two different methods were tracked by β-ray-induced X-ray spectrometry (BIXS). Exposure to tritium gas and irradiation of tritium ions were employed. Effects of pre-irradiation of helium ions on tritium retention were also investigated. After charging tritium by these methods and tracking by BIXS, thermal release behavior of tritium was examined by BIXS. The observed X-ray spectrum showed clearly that thermally charged tritium at a high temperature diffused into the bulk of SS316, but tritium ions irradiated at room temperature were trapped in surface layers of the sample irrespective of tritium ion energy. Thermally charged tritium was not able to remove completely even heating at high temperature of 973K. On the contrary, it was seen that the ion-irradiated tritium decreased almost linearly with temperature irrespective of ion energy. Furthermore, rapid decrease of tritium appeared in the helium pre-irradiated samples.

Keywords: Tritium retention, Tritium ion irradiation, Helium pre-irradiation, BIXS Stainless steel

Accepted: 12 August 2016

Vladimir Kh. Alimova^a,*, Yuji Hatano^a, Kazuyoshi Sugiyama^b, Sosuke Kondo^c,Tatsuya Hinoki^c, Masayuki Tokitani^d
aUniversity of Toyama, Toyama 930-8555, Japan
^bMax-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany
^cInstitute of Advanced Energy, Kyoto University, Uji 611-0011, Japan
^dNational Institute for Fusion Science, Toki 509-5292, Japanh

Abstract

The F82H and Eurofer97 steel samples were irradiated at 300 K with 20 MeV W ions to the damage level of 0.54 displacements per atom (dpa) at the damage peak. Additionally, the F82H steel samples were irradiated at 523 K with 6.4 MeV Fe ions to various damage levels in the range from 0.02 to 12.5 dpa. The damaged samples were exposed to D₂ gas at a pressure of 100 kPa and various temperatures in the range from 373 to 573 K for a certain length of time sufficient to fill ion-induced defects with deuterium. Trapping of deuterium at the ion-induced defects was examined by the D(³He, p)⁴He nuclear reaction with ³He energies between 0.69 and 4.0 MeV allowing determination of the D concentration up to a depth of 7 μm. It has been found that (i) at the damage level above 0.5 dpa, the concentration of the ion-induced defects responsible for trapping of diffusing D atoms does not depend practically on the numbers of displacements per atom, and (ii) the saturation value of the D concentration in the damage zone decreases with increasing D₂ gas exposure temperature, T_exp, and varies from about 10^-1 at.% at T_exp = 373 K to 10_-3 at.% at T_exp = 573 K. The deuterium-trap binding energy is estimated to be 0.7 ± 0.2 eV.

Keywords: Tritium retention, Tritium ion irradiation, Helium pre-irradiation, BIXS Stainless steel

Accepted: 12 August 2016

Masanori Hara¹, Masato Nakayama¹, Kiyokazu Hirokami², Tsukasa Aso^3
1Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama City, Toyama 930-8555, Japan
²Radioisotope Laboratory, Center for Basic Research and Development in Natural Sciences, University of Toyama, Gofuku 3190, Toyama City, Toyama 930-8555, Japan
³Department of Electronics and Computer Engineering, National Insutitute of Technology, Toyama College, Ebie-neriya 1-2, Imizu City, Toyama 933-0293, Japan

Abstract

To know the appropriate quenching level, the modified integral counting method with various quenched samples (MICM-VQ) was applied for the determination of the disintegration rate of ¹⁴C or ³⁵S. The appropriate quenching level for the MICM-VQ was considered by the comparison of the integral scintillation spectrum and the integral beta spectrum. The appropriate quenching level of sample in the MICM-VQ was found. The disintegration rate of sample having appropriate quenching level could be determined by the MICM-VQ.

Keywords: Beta emitters, Integral scintillation spectrum, Beta spectrum, Disintegration rate, Modified integral counting method

Published online: 28 April 2016

Yuki Iwasa^a,∗, Kohei Yamanoi^a, Keisuke Iwano^a, Melvin John F. Empizo^a,Yasunobu Arikawa^a, Shinsuke Fujioka^a, Nobuhiko Sarukura^a, Hiroyuki Shiraga^a, Masaru Takagi^a, Takayoshi Norimatsu^a, Hiroshi Azechi^a, Kazuyuki Noborio^b, Masanori Hara^b, Masao Matsuyama^b
aInstitute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871, Japan
^bHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract

We investigate the tritium-doping enhancement of polystyrene by ultraviolet (UV) laser and hydro-gen plasma irradiation. Tritium-doped polystyrene films are fabricated by the Wilzbach method withUV laser and hydrogen plasma. The 266-nm laser-irradiated, 355-nm laser-irradiated, and hydrogenplasma-irradiated polystyrene films exhibit higher PSL intensities and specific radioactivities than thenon-irradiated sample. Tritium doping by UV laser irradiation can be largely affected by the laserwavelength because of polystyrene’s absorption. In addition, UV laser irradiation is more localized andconcentrated at the spot of laser irradiation, while hydrogen plasma irradiation results to a more uniformdoping concentration even at low partial pressure and short irradiation time. Both UV laser and plasmairradiations can nevertheless be utilized to fabricate tritium-doped polystyrene targets for future laserfusion experiments. With a high doping rate and efficiency, a 1% tritium-doped polystyrene shell targethaving 7.6 × 10¹¹ Bq g^-1 specific radioactivity can be obtained at a short period of time thereby decreasingtritium consumption and safety management costs.

Keywords: Tritium, Polystyrene, Laser fusion, Wilzbach method, Irradiation

Accepted: 7 September 2016

Y. Hatano^a,∗, K. Ami^a, V.Kh. Alimov^a , S. Kondo^b, T. Hinoki^b, T. Toyama^c, M. Fukuda^d, A. Hasegawa^d, K. Sugiyama^e, Y. Oya^f, M. Oyaidzu^g, T. Hayashi^g
aHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan
^bInstitute of Advanced Energy, Kyoto University, Uji 611-0011, Japan
^cIntitute for Materials Research, Tohoku University, Oarai 311-1313, Japan
^dSchool of Engineering, Tohoku University, Sendai 980-8579, Japan
^eMax-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany
^fCollege of Science, Academic Institute, Shizuoka University, Shizuoka 422-8529, Japan
^gSector of Fusion Research and Development, Japan Atomic Energy Agency, Rokkasho 039-3212, Japan

Abstract

Neutron irradiation to W induces defects acting as traps against hydrogen isotopes and transmutation elements such as Re and Os. To investigate synergetic effects on radiation-induced defects and Re, deuterium (D) retention in W and W–5 Re samples were examined after irradiation with 6.4 MeV Fe ions at 523–1273 K followed by exposure to D2 gas at 673 K. The value of D retention in W–5% Re was lower than that in W by orders of magnitude after the irradiation at high temperatures (≥1073 K), while no significant effects of Re addition was observed after i rradiation at 523 K. Irradiation with 20 MeV W ions at room temperature followed by exposure to D plasma at 443–743 K also resulted in small difference in D retention between W and W–5% Re samples. The results of positron lifetime measurements showed that the reduced D retention by Re observed after high temperature irradiation was due to suppression of formation of vacancy-type defects (monovacancies and vacancy clusters) by Re.

Keywords: Tungsten, Tungsten-rhenium alloy, Irradiation, Defect, Trap, Hydrogen, Hydrogen isotope

Accepted: 30 June 2016

Yuji Nobuta^a, Yuji Hatano^b, Yuji Torikai^b, Masao Matsuyama^b, Shinsuke Abe^b, Yuji Yamauchi^a,
^aLaboratory of Plasma Physics and Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628, Japan
^bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

In this study, 1 keV DT⁺ion irradiation was performed on tungsten pre-irradiated with helium. The ther-mal desorption behavior and the reduction of tritium retention during vacuum preservation at roomtemperature, as well as isochronal annealing were investigated using an IP technique, taking advantageof the fact that tritium detection is nondestructive and is highly sensitive. At a pre-irradiated helium flu-ence of 1 × 10 ¹⁷ He/cm², retained tritium tended to be desorbed at higher temperatures when comparedto no helium irradiation case. Tritium retention during preservation in vacuum and during isochronalannealing became smaller with increasing helium fluence up to 1 × 10¹⁷He/cm2. At a helium fluence of 1 × 10¹⁸He/cm², the reduction of tritium retention was found to be greater compared to 1 × 10¹⁷He/cm².The results indicate that helium irradiation has a significant influence not only on the thermal desorptiontemperature of tritium but on longtime tritium reduction at room and elevated temperatures.

Keywords: Tritium retention, Tungsten, Ion irradiation, Thermal desorption, Isochronal heating

Accepted: 23 December 2015

Cui Hu^a, Yuki Uemura^a, Kenta Yuyama^a, Hiroe Fujita^a, Shodai Sakurada^a, Keisuke Azuma^a, Akira Taguchi^b, Masanori Hara^b, Yuji Hatano^b, Takumi Chikada^a, Yasuhisa Oya^a,
^aShizuoka University, 836 Ohya, Suruga-ku Shizuoka 422-8529, Japan
^bUniversity of Toyama, 3190 Gofuku, Toyama 939-8555, Japan

Abstract

Usage of CuO and water bubbler is one of the conventional and convenient methods for tritium recovery. In present work, influence of CuO particle size and diameter of reaction tubing on the tritium recoverywas evaluated. Reaction rate constant of tritium with CuO particle has been calculated by the combina-tion of experimental results and a simulation code. Then, these results were applied for exploring thedependence of reaction tubing length on tritium conversion ratio. The results showed that the surfacearea of CuO has a great influence on the oxidation rate constant. The frequency factor of the reactionwould be approximately doubled by reducing the CuO particle size from 1.0 mm to 0.2 mm. Cross sectionof reaction tubing mainly affected on the duration of tritium at the temperature below 600 K. Reactiontubing with length of 1 m at temperature of 600 K would be suitable for keeping the tritium conversionratio above 99.9 %. The length of reaction tubing can be reduced by using the smaller CuO particle orincreasing the CuO temperature.

Keywords: Tritium recovery, CuO, Water bubbler, Particle size, Reaction tubing

Accepted: 13 May 2016

Y. Oya^a, H. Fujita^a, C. Hu^a, Y. Uemura^a, S. Sakurada^a, K. Yuyama^a, X. Li^a, Y. Hatano^b, N. Yoshida^c, H. Watanabe^c, Y. Nobuta^d, Y. Yamauchi^d, M. Tokitani^{e, f}, S. Masuzaki^e, T. Chikada^a
aGraduate School of Science, Shizuoka University, Shizuoka, Japan
^bHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama, Japan
^cInstitute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
^dGraduate School of Engineering, Hokkaido University, Sapporo, Japan
^eDepartment of Helical Plasma Research, National Institute for Fusion Science, Toki, Japan
^fSOKENDAI (Grad. University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292, Japan

Abstract

Effect of carbon based mixed-material deposition layer formation on hydrogen isotope retention was studied. The tungsten (W) samples were placed at four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (higher heat load area), and ER (erosion dominated area) during 2013 plasma experimental campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ~4000 shots of hydrogen plasma in a 2013 plasma experimental campaign. Most of the sample surface except for ER was covered by a mixed-material de- position layer formed by plasma experimental campaign, which consisted of carbon, but some metal im-purities were contained. For ER sample, He bubbles were formed due to long term He discharge cleaning and He plasma experiments during the plasma experimental campaign. The additional 1 keV D₂⁺ implan- tation was performed to evaluated the D retention enhancement by plasma exposure. It was found that both of H and D retentions were clearly increased. In particular, the H retention was controlled by the thickness of the carbon-dominated mixed-material deposition layer, indicating most of the H was trapped by this mixed-material deposition layer. It is concluded that the accumulation of low-Z mixed-material layer on the surface of the first wall is one of key issues for the determination of hydrogen isotope re- tention in first wall.

Keywords: Hydrogen isotope retention enhancement, Mixed-material layer, TEM, TDS, XPS, LHD

Accepted: 1 July 2016

K. Katayama^a, Y. Someya^b, K. Tobita^b, S. Fukada^a, Y. Hatano^c, T. Chikada^d
aDepartment of Advanced Energy Engineering Science, Kyushu University 6-1, Kasugakoen, Kasuga-shi, Fukuoka 816-8580, Japan
^bNational Institutes for Quantum and radiological Science and Technology, 2-166 Omotedate, Obuchi, Rokkasho-mura, Kamikita-gun, Aomori 039-3212, Japan
^cHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^dDepartment of Chemistry, Graduate school of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529, Japan

Abstract

Hydrogen addition to a sweep gas of a solid breeder blanket module has been proposed to enhancetritium recovery from the surface of the breeders. However, the influence of hydrogen addition on thebred tritium behavior is not understood completely. Tritium behavior in the simplified blanket moduleof Li₂ TiO₃ pebbles was numerically calculated considering diffusion in the grain bulk, surface reactionson the grain surface and permeation through the cooling pipe. Although a partial pressure of T₂ increaseswith increasing a partial pressure of H₂ in the sweep gas, it was estimated that tritium permeation rate tothe cooling water decreases. Additionally, the release duration of water vapor generated by the reactionof the pebbles and hydrogen is shortened with increasing a partial pressure of H₂. Tritium inventory inthe grain bulk and the grain surface occupies 99.6 % of total tritium inventory in the blanket module.

Keywords: Tritium, Li₂TiO₃, Permeation, Isotope exchange, Water formation

Accepted: 20 August 2016

Satoshi Fukada^a, Yasuhisa Oya^b, Yuji Hatano^c
aDept. Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-Koen, Kasuga, 816-8580, Japan
^bCollege of Science, Academic Institute, Shizuoka University, 836 Otani, Suruga-ku, Shizuoka 422-8529, Japan
^cHydrogen Isotope Research Center, Organization for Promotion Research, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract

After introduction of Japanese history or recent topics on tritium (T)-relating research and T-handlingcapacity in facilities or universities, present activities on T engineering research in Japan are summa-rized in short in terms of T accountability on safety. The term of safety includes wide processes fromT production, assay, storing, confinement, transfer through safety handling finally to shipment of itswaste. In order to achieve reliable operation of fusion reactors, several unit processes included in the Tcycle of fusion reactors are investigated. Especially, the following recent advances are focused: T reten-tion in plasma facing materials, emergency detritiation system including fire case, T leak through metaltube walls, oxide coating and water detritiation. Strict control, storing and accurate measurement areespecially demanded for T accountability depending on various molecular species. Since kg-order T ofvaporable radioisotope (RI) will be handled in a fuel cycle or breeding system of a fusion reactor, the accu-racy of <0.1 % is demanded far over the conventional technology status. Necessity to control T balancewithin legal restrictions is always kept in mind for operation of the future reactor.

Keywords: Tritium, Review, Accountability, Safety, Self-sufficiency

Accepted: 22 August 2016

Mitsuhiro Inoue ^a, Yuta Takahashi ^b, Mitsuhiko Katagiri ^b, Takayuki Abe ^a,Minoru Umeda ^{b, *
a} Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^b Department of Materials Science and Technology, Faculty of Engineering, Nagaoka University of Technology, Kamitomioka 1603-1, Nagaoka, Niigata 940-2188, Japan

Abstract
The surface modification of Al₂O₃ particles with Au was investigated using the polygonal barrelsputtering method. When the Au was sputtered, a hexagonal barrel loaded with Al₂O₃ particles was oscillated. As a result, the appearance of the sample changed from white to dark brown. The characterization of the prepared sample showed that the surfaces of the Al₂O₃ particles were uniformly modified with Au having a face-centered cubic structure. The electrical conductivity of the prepared sample was very high (σ=1.19 × 10³-2.45 ×10³ S m^-1) and its electrochemical property was similar to that of the bulk Au metal. Thus, the polygonal barrel-sputtering method is concluded to be useful for providing an electrical conductivity to insulators such as metal oxide particles.

Yasuhisa Oya^a,∗, Yuji Hatano^b, Masashi Shimada^c, Dean Buchenauer^d, Robert Kolasinski^d, Brad Merrill^c, Sosuke Kondo^e, Tatsuya Hinoki^e, Vladimir Kh. Alimov^b
aShizuoka University
^bUniversity of Toyama
^cIdaho National Laboratory, Idaho Falls, ID 83415, USA
^dSandia National Laboratories, Livermore, CA 94551, USA
^eKyoto University

Abstract
This paper reviews recent results pertaining to hydrogen isotope behavior in neutron and heavy iondamaged W. Accumulation of damage in W creates stable trapping sites for hydrogen isotopes, therebychanging the observed desorption behavior. In particular, the desorption temperature shifts higher asthe defect concentration increases. In addition, the distribution of defects throughout the sample alsochanges the shape of TDS spectrum. Even if low energy traps were distributed in the bulk region, theD diffusion toward the surface requires additional time for trapping/detrapping during surface-to-bulktransport, contributing to a shift of desorption peaks toward higher temperatures. It can be said that bothof distribution of damage (e.g. hydrogen isotope trapping sites) and their stabilities would have a largeimpact on desorption. In addition, transmutation effects should be also considered for an actual fusionenvironment. Experimental results show that production of Re by nuclear reaction of W with neutronsreduces the density of trapping sites, though no remarkable retention enhancement is observed.

Keywords:Hydrogen isotope behavior in damaged W, Neutron irradiation, Heavy ion irradiation, Plasma wall interactions
Accepted:5 August 2016

Masami Ohnishi^a,∗, Yasushi Yamamoto^a, Hodaka Osawa^a, Yuji Hatano^b, Yuji Torikai^b,Isao Murata^c, Keita Kamakura^a, Masaaki Onishi^a, Keiji Miyamoto^a, Hiroki Konda^a,Kai Masuda^d, Eiki Hotta^e
aDepartment of Science and Engineering, Kansai University
^bHydrogen Isotope Science Center, University of Toyama
^cFaculty of Engineering Environment and Energy Department, Osaka University
^dInstitute of Advanced Energy, Kyoto University
^eInterdisciplinary Graduate School of Science and Engineering

Abstract
An experiment on tritium burning is conducted to investigate the enhancement in the neutron productionrate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it isshielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuumchamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritiumis used, and its neutron production rate is measured to be 5–8 times more than that of pure deuteriumgas. Moreover, the results show good agreement with those of a simplified theoretical estimation ofthe neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recoveryprocedure through a water bubbler device. The amount of gaseous tritium released by the developedIECF facility after tritium burning is verified to be much less than the threshold set by regulations.

Keywords:D–T burning, Neutron production, Tritium safety, Getter pump, Water bubbler, Tritium recovery
Accepted: 22 October 2015

Yuji Hatano

Abstract
Because of their small sizes in molecular and atomic forms, hydrogen isotopes easily dissolve in a solid material and permeate through it. The permeation through a container material is a critical issue against safe confinement of tritium. On the other hand, it is common to employ a permeation technique for the separation of hydrogen isotopes from other gaseous species. In this chapter, elemental processes of permeation of hydrogen isotopes through metals and ceramics are explained together with the possible isotope effects on them. Fundamental equations describing the hydrogen permeation through materials under the exposure to gas and plasma are given. The hydrogen permeation under the corrosive environment is also discussed. Characteristics of hydrogen dissolution, diffusion , and permeation in metallic and ceramic materials important for fusion are summarized together with the peculiarity of the hydrogen permeation under the fusion reactor environments.

Keywords: Tritium, Hydrogen isotopes, Permeation, Surface, Diffusion
DOI: 10.1007/978-4-431-56460-7_9