発表論文 2021

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[2021_01]

Iron catalysts supported on nitrogen functionalized carbon for improved CO2 hydrogenation performance

Rungtiwa Kosol a, Lisheng Guo a,*, Naoya Kodama a, Peipei Zhang a,*, Prasert Reubroycharoen b, Tharapong Vitidsant b, Akira Taguchi c, Takayuki Abe c, Jienan Chen d, Guohui Yang a, Yoshiharu Yoneyama a, Noritatsu Tsubaki a,*

a Department of Applied Chemistry, School of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
b Department of Chemical Technology, Faculty of Science, Chulalongkorn University, Bangkok 10330, Thailand
c Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
d Ministry of Forestry Bioethanol Research Center, Central South University of Forestry and Technology, Changsha 410004, China

Abstract
Preparation of highly efficient Fe-based catalysts is a reliable and achievable goal for catalyzing CO2 hydrogenation. Herein, ethylene diamine as a benign modifier well regulates the surface properties of carbon support, achieving a good dispersion of active small-size iron carbide sites. With the further incorporation of alkaline K promoter, heavy hydrocarbon selectivity (C5+) is increased from 14.8% to 39.8%. Combining several catalyst characterization (XRD, CO2-TPD, H2-TPR, TEM, and XPS) and reaction data, discloses that good dispersion, enhanced reduction/carburization behavior, and small-size carbides formation are essential for improving CO2 performance. Simultaneous doping of nitrogen atoms and alkali metal provides a promising means for CO2 fixation and rational design of functionalized metal-supported carbon catalysts.

Keywords:Nitrogen incorporation, CO2hydrogenation, Carbon materials, Fe-based catalyst, Alkaline promoter
Accepted:29 October 2020

[2021_02]

Effects of fabrication conditions on the microstructure, pore characteristics and gas retention of pure tungsten prepared by laser powder bed fusion

Takafumi Yamamoto a,b,*, Masanori Hara b, Yuji Hatano b
a Toyama Industrial Technology Research and Development Center
b University of Toyama, Organization for Research Promotion, Hydrogen Isotope Research Center

Abstract
In this study, the effects of fabrication conditions on the microstructure, pore characteristics and gas retention of pure tungsten specimens prepared by laser powder bed fusion (LPBF) were investigated. The LPBF specimens contained micro- and nano- sized pores as internal defects. By optimizing the laser irradiation conditions, the formation of micropores was suppressed. The densest LPBF specimen was obtained when the input energy density was adjusted to be 411 J/mm3, and the relative density of the specimen measured by utilizing Archimedes’principle reached 98.58 ± 0.25% (19.03 ± 0.05 g/cm3). Heat treatment at a high temperature (1900 ℃) was effective to reduce the number density of nanopores. The compositions and amounts of internal gas in the specimens were examined using a quadrupole mass spectrometer. The results indicated that the specimens contained the argon (Ar), which was used as the shielding gas. The Ar retention was correlated with the number density of nanopores in the specimens and not with the density of micropores. The Ar retention decreased to almost half that in the as-fabricated specimen after the heat treatment at 1900 ℃. These observations suggest that Ar gas was trapped in nanopores in the LPBF material. It was demonstrated that a part of the trapped Ar dissolved and diffused in grains during the heat treatment and was released to the outside.

Keywords:Additive manufacturing,Laser powder bed fusion,Pure tungsten,Microstructure,Gas retention
Accepted :14 October 2020

[2021_03]

Effects of Helium Seeding on Deuterium Retention in Neutron-Irradiated Tungsten

Yuji Nobuta,a* Masashi Shimada,b Chase N. Taylor,b Yasuhisa Oya,c Yuji Hatano,d Yaqiao Wu,e,f and Megha Dubeye,f
aHokkaido University, Sapporo, Japan
bIdaho National Laboratory, Idaho Falls, Idaho
cShizuoka University, Shizuoka, Japan
dUniversity of Toyama, Toyama, Japan
eBoise State University, Boise, Idaho
fCenter for Advanced Energy Studies, Idaho Falls, Idaho

Abstract
— Neutron -irradiated tungsten (W) samples were exposed to helium (He)–seeded deuterium (D) plasmas using a linear plasma device called Tritium Plasma Experiment in order to investigate the synergetic effects of neutron and He irradiations on D retention in W. Exposure to nonseeded D plasma was also performed for neutron-irradiated and nonirradiated W samples for comparison. Deuterium retention in neutron-irradiated W after D plasma exposure was two to three times larger than that in W without neutron irradiation. Nevertheless, He seeding in D plasma resulted in a drastic reduction in D retention. The cross-sectional observation by transmission microscopy showed formation of He bubble layers with a thickness of 10 to 20 nm. There is a possibility that alpha particles in fusion plasma reduce tritium retention in neutron-irradiated plasma-facing components with W layers.

Keywords:Tungsten, neutron irradiation, hydrogen isotope retention, helium bubble.
Accepted:October 26, 2020

[2021_04]

Tritium retention in displacement-damaged tungsten exposed to deuterium-tritium gas mixture at elevated temperatures

V.Kh. Alimov a,b,c,*, Y. Torikai a,1, Y. Hatano a, T. Schwarz-Selinger d

a Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama, 930-8555, Japan
b A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow, 119071, Russia
c National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, 115409, Russia
d Max-Planck-Institut für Plasmaphysik, D-85748, Garching, Germany

Abstract
W samples, previously irradiated with 20 MeV W ions at room temperature to a displacement-damage level of 0.23 displacements per atom (dpa) at the peak of displacements, were exposed to a deuterium-tritium (D2 – DT –T2) gas mixture with a tritium content of 6% at a total pressure of 1.2 kPa and temperatures of 773 K and 973 K for 3 h. The concentration of tritium retained in the displacement-damage zone of these W samples was determined by a method combining chemical etching and subsequent analysis of tritium in the etching solution using liquid scintillation counting (CE-LSC). (Some figures may appear in colour only in the online journal)

Keywords:Tungsten,Ion-induced defects,Deuterium-tritium gas mixture Liquid scintillation counting
Accepted:28 October 2020

[2021_05]

Deuterium retention in W and binary W alloys irradiated with high energy Fe ions

Jing Wang a , Yuji Hatano a , ∗, Tatsuya Hinoki b, Vladimir Kh. Alimov c,d, Alexander V. Spitsyn c, Nikolay P. Bobyr c, Sosuke Kondo e, Takeshi Toyama e, Heun Tae Lee f, Yoshio Ueda f, Thomas Schwarz-Selinger g

aHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan
b Institute of Advanced Energy, Kyoto University, Kyoto 611-0011, Japan
c NRC “Kurchatov Institute”, Kurchatov sq. 1, Moscow 123182, Russia
d A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russia
eInstitute for Materials Research, Tohoku University, Miyagi 980-8577, Japan f Graduate School of Engineering, Osaka University, Osaka 565-0871, Japan
g Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, Garching D-85748, Germany

Abstract
To investigate systematically the effects of Re and other elements on deuterium (D) retention, W and bi- nary W-alloys (Re, Mo and Ta) were irradiated with 6.4 MeV Fe ions at high temperatures (1073–1273 K) and then exposed to D 2 gas at 673 K. Depth profiles of D were measured by nuclear reaction analysis (NRA), and D retention was determined by thermal desorption spectrometry (TDS) and NRA. The addition of 5 at.% Re into W reduced the content of D trapped at radiation-induced defects created by irradiation with the Fe ions at 1273 K to the peak damage level of 5 displacements per atom (dpa). At Re fractions of 1, 3 and 5 at.%, comparable effects on D retention were observed after irradiation to 0.5 dpa at 1073 K. The addition of Mo and Ta to W resulted in no visible effects in D retention.

Keywords:Tungsten,Rhenium, Molybdenum, Tantalum, Deuterium retention, Ion irradiation
Accepted:15 December 2020

[2021_06]

Global distribution of tritium in JET with the ITER-like wall

S.E. Lee a, Y. Hatano a,*, M. Tokitani b, S. Masuzaki b, Y. Oya c, T. Otsuka d, N. Ashikawa b, Y. Torikai e, N. Asakura f, H. Nakamura f, K. Isobe f, H. Kurotaki f, D. Hamaguchi f, T. Hayashi f, A. Widdowson g, S. Jachmich h, J. Likonen i, M. Rubel j, JET contributors1

a University of Toyama, Toyama 930-8555, Japan
b National Institute for Fusion Science, 322-6, Toki 509-5292, Japan
c Shizuoka University, Oya 836, Shizuoka 422-8529, Japan
d Kindai University, Kowakae 3-4-1, Higashiosaka 577-8502, Japan
e Ibaraki University, Bunkyo 2-1-1, Mito 310-8512, Japan
f National Institutes for Quantum and Radiological Science and Technology, Rokkasho 039-3212, Japan
g Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, UK
h Association Euratom-Etat Belge, ERM-KMS, Brussels Belgium
i VTT, Otakaari 3J, P.O. Box 1000, FIN-02044, VTT, Finland
j KTH Royal Institute of Technology, 100 44 Stockholm, Sweden

Abstract
Nondestructive analysis of tritium (T) distribution was performed by means of imaging plate technique on specimens cut from the Be limiters, W-coated carbon tiles and bulk W lamellae retrieved from the JET tokamak after the first and third experimental campaigns with the ITER-like wall. Afterwards, analyses were continued using X-ray photoelectron spectroscopy, microscopy techniques and thermal desorption spectroscopy. Co-deposits formed on the W-coated tiles in the 1st campaign showed large T retention because of high carbon content reaching up to 50 atomic %, while the carbon fraction in co-deposits after the 3rd campaign was distinctly lower. The T retention of the plasma-facing surface of the bulk W tile was smaller than that of the W-coated tiles by a factor of 20, while deposition of small amount of T was found at the side surfaces facing to the gaps in a lamella structure. The correlation of T distributions with surface morphology and the discharge conditions is discussed.

Accepted:26 January 2021

[2021_07]

Effects of irradiation temperature on tritium retention in stainless steel type 316L

Masao Matsuyama a,*, Hideki Zushi b, Kazuaki Hanada b, Yasuhisa Oya c, Yuji Hatano a
a Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama
b Research Institute for Applied Mechanics, Kyushu University
c Graduate School of Science and Technology, Shizuoka University

Abstract
Dependence of irradiation temperature on tritium retention has been studied using stainless steel type 316L (SS316L) as a model sample. A mixture of D2+ and DT+ ions was used and two kinds of ion energy, 0.5 and 2.5 keV, were applied. In case of irradiation tests by 0.5 keV, tritium retention decreased with increasing temperature up to 523 K, while above this temperature it contrarily showed an increase tendency. Such a concave change was not observed for irradiation tests at 2.5 keV. The retention was almost same until 400 K, but above this temperature it decreased gradually. It was seen from the analyses by X-ray photoelectron spectroscopy that most of surface was initially covered with the carbon and oxygen species at room temperature. Among metallic elements, constituents such as Fe and Ni were metallic states more than 60 % at room temperature, while metallic chromium atoms were little observed. Both fractions of the metallic chromium and iron atoms in the major base metals of SS316 L increased with an increase in temperature, but metallic nickel atoms relatively decreased. It was suggested, therefore, that real surface states of the irradiation materials play an important role for behavior of tritium retention.

Keywords:Tritium retention, Irradiation temperature, SS316L, BIXS, XPS
Accepted:1 March 2021

[2021_08]

Deuterium retention in tungsten irradiated by high-dose neutrons at high temperature

M. Oya a,*, M. Shimada b, C.N. Taylor b, M.I. Kobayashi c, Y. Nobuta d, Y. Yamauchi d, Y. Oya e,Y. Ueda f, Y. Hatano g

a Faculty of Engineering Sciences, Kyushu University
b Fusion Safety Program, Idaho National Laboratory
c National Institute for Fusion Science
d Graduate School of Engineering, Hokkaido University
e College of Science, Academic Institute, Shizuoka University
f Graduate School of Engineering, Osaka University
g Hydrogen Isotope Research Center, University of Toyama

Abstract
We investigated deuterium (D) retention in three W samples irradiated with MeV neutrons at high damage level of 0.39 ~ 0.74 displacements per atom (dpa) at high temperatures, 894 K, 1074 K and 1379 K. The W specimens were exposed to high-flux (~1 × 1022 m -2 s -1) and high-fluence (~5 × 1025 m-2) D plasma at 873 K in the Tritium Plasma Experiment. Broad desorption peaks extended from 900 K to 1200 K were observed for the neutron-irradiated W by thermal desorption spectroscopy (TDS). The retention in neutron-irradiated specimens was much larger than for an un-irradiated specimen. The highest D retention was obtained for a specimen irradiated at 894 K. With increasing neutron irradiation temperature, the retention was reduced about by half at 1074 K and further increase of the temperature (1379 K) resulted in comparable retention. In addition, onedimensional diffusion calculations (D desorption in TDS and D depth distribution in plasma exposure) were performed to derive retention parameters (the detrapping energy, the depth occupied by D atoms and D/W ratio) from experimental D retention properties of neutron-irradiated W. By TDS simulation calculation, simple dependences of the peak temperature, height and width of TDS peaks on the retention parameters were obtained with total retention in the orders of 1019 ~ 1022 m-2. The calculation of the depth distribution of trapped D atoms made a relationship between the D/W ratio and the depth occupied by D atoms after plasma exposure at relevant conditions. By comparing the relationship (the D/W and the depth) with that obtained from the experimental results, we estimate each retention parameters for the specimens irradiated by high-dose neutrons at the high temperatures. And, we discussed the neutron-irradiation temperature dependence of D retentions.

Keywords:Tungsten, Deuterium retention, Neutron irradiation, Plasma-material interaction
Accepted:16 March 2021

[2021_09]

Conceptual Design of HFIR Irradiation Experiment for Material Compatibility Study on Liquid Sn Divertor

Masatoshi KONDO, Bruce A. PINT1), Jiheon JUN1), Nick RUSSELL1), Joel McDUFFEE1), Masafumi AKIYOSHI2), Teruya TANAKA3), Naoko OONO4), Junichi MIYAZAWA3), Josina W GERINGER1), Yutai KATOH1) and Yuji HATANO5) Tokyo Institute of Technology
1)Oak Ridge National Laboratory, Oak Ridge
2)Osaka Prefecture University
3)National Institute for Fusion Science
4)Hokkaido University
5)University of Toyama

Abstract
Liquid Sn is one of the promising coolants for liquid surface divertor concept of fusion reactors. However, the compatibility between liquid Sn and structural materials is an important issue that has to be addressed, because liquid Sn is extremely corrosive to steels at high temperatures. The corrosion may be mitigated when a protective Al2O3 layer is formed on the surface of alumina forming steels. However, the effect of neutron irradiation on the integrity of protective layer is not made clear so far. Japan and US joint research project “FRONTIER” started in 2019 to investigate the material compatibility under neutron irradiation. The purpose of the present study is to develop the conceptual design of the irradiation test capsule which enables material compatibility tests for the alumina forming steels – liquid metal systems under neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory, TN, USA. The three dimensional drawing of capsule structure was then developed. The validity of the material selections for the capsule design was investigated by means of corrosion tests of SiC, Si3N4, Ti, and Mo in liquid Sn at 773K for 262 hr.

Keywords: liquid metal, irradiation, divertor, compatibility, irradiation test capsule
Accepted: Accepted 8 February 2021

[2021_10]

Effects of fabrication conditions on the microstructure, pore haracteristics and gas retention of pure tungsten prepared by laser powder bed fusion

Takafumi Yamamoto a,b,*, Masanori Hara b, Yuji Hatano b
a Toyama Industrial Technology Research and Development Center
b University of Toyama, Organization for Research Promotion, Hydrogen Isotope Research Center

Abstract
In this study, the effects of fabrication conditions on the microstructure, pore characteristics and gas retention of pure tungsten specimens prepared by laser powder bed fusion (LPBF) were investigated. The LPBF specimens contained micro- and nano- sized pores as internal defects. By optimizing the laser irradiation conditions, the formation of micropores was suppressed. The densest LPBF specimen was obtained when the input energy density was adjusted to be 411 J/mm3, and the relative density of the specimen measured by utilizing Archimedes’ principle reached 98.58 ± 0.25% (19.03 ± 0.05 g/cm3). Heat treatment at a high temperature (1900 ℃) was effective to reduce the number density of nanopores. The compositions and amounts of internal gas in the specimens were examined using a quadrupole mass spectrometer. The results indicated that the specimens contained the argon (Ar), which was used as the shielding gas. The Ar retention was correlated with the number density of nanopores in the specimens and not with the density of micropores. The Ar retention decreased to almost half that in the as-fabricated specimen after the heat treatment at 1900 ◦C. These observations suggest that Ar gas was trapped in nanopores in the LPBF material. It was demonstrated that a part of the trapped Ar dissolved and diffused in grains during the heat treatment and was released to the outside.

Keywords:Additive manufacturing,Laser powder bed fusion,Pure tungsten,Microstructure,Gas retention
Accepted:14 October 2020

[2021_11]

Numerical analysis of deuterium migration behaviors in tungsten damaged by fast neutron by means of gas absorption method

Makoto I Kobayashi a,b,*, Masashi Shimada c, Chase N Taylor c, Yuji Nobuta d, Yuji Hatano e, Yasuhisa Oya f
a National Institute for Fusion Science
b The Graduate University for Advanced Studies
c Idaho National Laboratory
d Hokkaido University
e Hydrogen Isotope Research Center, University of Toyama f Shizuoka University

Abstract
Deuterium retention behavior in tungsten damaged by fast neutrons at high temperatures (0.43 dpa at 918 K and 0.74 dpa at 1079 K) and 6.4 MeV Fe2+ (0.3 dpa at R.T.) were nvestigated to evaluate the tritium retention property of fusion reactor divertors. A deuterium gas absorption method was carried out to avoid additional damage that may be induced by plasma exposure, then, deuterium retention and desorption behaviors were investigated quantitatively by means of thermal desorption spectroscopy and the following simulation code. The deuterium desorption spectra for tungsten samples were analyzed by the numerical code which includes the elementary steps of hydrogen isotope migration processes including diffusion, trapping, detrapping, and surface recombination. The evaluated deuterium detrapping energy from the irradiation defects in neutron irradiated tungsten sample was larger than that in 6.4 MeV Fe2+ irradiated tungsten. It was suggested that the dominant deuterium trapping site in the neutron irradiated tungsten would be voids which was formed by the accumulation of vacancies during neutron irradiation under high temperature and long duration.

Keywords:Tungsten, Neutron, Divertor, TDS
Accepted:26 April 2021

[2021_12]

Protective behavior of tea catechins against DNA double strand breaks produced by radiations with different linear energy transfer

Takuro Wada a, Ayaka Koike a, Shota Yamazaki a, Kyosuke Ashizawa a, Fei Sun b, Yuji Hatano c, Hiroto Shimoyachi d, Takahiro Kenmotsu e, Takashi Ikka f,g, Yasuhisa Oya a,*

a Graduate School of Integrated Science and Technology
b Faculty of Science
c Hydrogen Isotope Research Center, University of Toyama
d Graduate School of Science and Engineering
e Faculty of Life and Medical Sciences, Doshisha University
f Faculty of Agriculture, Shizuoka University
g Institute for Tea Science, Shizuoka University

Abstract
several types of radiation sources. The characteristics of radiation damages by different linear energy transfer (LET) and the effect of radiation protection by tea catechins were also studied as a function of its concentration. After β-rays and γ-rays irradiation to samples, the length of genome size DNA molecules (bacteriophage T4 GT7 DNA; 166 kbp) was measured by single molecule observation method using fluorescence microscope, which can estimate DSBs quantitatively. It was found that the number of DSBs was increased with increasing LET due to high radical density. By addition of EGCg ((-)-epigallocatechin gallate), the number of DSBs was reduced with a small concentration of 1 μM.

Keywords:DNA damage, Tritium, Fluorescence microscope
Accepted:25 May 2021

[2021_13]

Effect of temperature distribution on tritium permeation rate to cooling water in JA DEMO condition

Kazunari Katayama a,*, Youji Someya b, Takumi Chikada c, Kenji Tobita d, Hirofumi Nakamura b,Yuji Hatano e, Ryoji Hiwatari b, Yoshiteru Sakamoto b, Akito Ipponsugi a, Makoto Oya a

a Department of Advanced Energy Engineering Science, Kyushu University
b National Institutes for Quantum and Radiological Science and Technology
c Graduate School of Science, Shizuoka University
d Department of Quantum Science and Energy Engineering, Tohoku University
e Hydrogen Isotope Research Center, University of Toyama

Abstract
The estimation of tritium permeation rate through the plasma facing wall into coolant is required to discuss tritium balance in a D-T fusion plant, to design tritium recovery system and to perform safety assessments. In this work, tritium permeation rates in the blanket first wall and the divertor were estimated by numerical analysis for simplified multi-layer structures with considering the temperature distribution in recent JA DEMO condition. The permeation rate in the blanket first wall, which was a double layer consisting of tungsten and F82H, was estimated to be 0.69 g/day. The permeation rate in the divertor, which was a triple layer consisting of tungsten,copper and copper alloy or F82H, was estimated to be 0.013 g/day. When the permeation rate in tritium breeding region in the blanket can be reduced by three orders of magnitude due to a permeation barrier, total tritium permeation rate in the blanket and the divertor was estimated to be 0.71 g/day.

Keywords:Tritium, Permeation, Blanket, Divertor, DEMO
Accepted:7 April 2021

[2021_14]

Working environment of tritium analysis for photoluminescence control

Mayu Ohki a, Tomomune Matsunaga a, Takuyo Yasumatsu a,*, Masanori Hara b
a Tokyo Power Technology Ltd.
b University of Toyama

Abstract
The conventional tritium analysis process using a liquid scintillation counter (LSC) requires over 24 h of waiting time to reduce the interference from the luminescence of the sample cocktail. The working efficiency of tritium measurement using an LSC worsens as the waiting time for luminescence decay increases. We hypothesize that this waiting time can be shortened by using light-emitting diode (LED) lamps as lighting equipment in a measurement laboratory because the emission spectra of some LED lamps contain no ultraviolet rays. Thus, a sample cocktail was prepared under an LED lamp. The count rate of this cocktail was reduced to the background of the LSC (ca. 3 cpm) within several hours. By contrast, the luminescence of a cocktail placed under daylight took approximately 100 h to decay. Therefore, the use of LED lamps is effective for luminescence control and shortening the waiting time in a measurement process.

Keywords:Liquid scintillation counter, Photoluminescence, LED lamp, Environmental sample, Tritium
Accepted:10 May 2021

[2021_15]

Monte Carlo simulation of the beta-ray induced X-ray spectra of tritium at various depths in solids

Masanori Hara a,*, Tomoya Shimura a, Kenjiro Aoki a, Masao Matsuyama a, Tsukasa Aso b, Marco Rölligc
a University of Toyama
b National Institute of Technology, Toyama College
c Karlsruhe Institute of Technology

Abstract
Non-destructive methods to measure tritium in solids are required for material studies and managements of tritium-contaminated material in the field of fusion research. The beta particle generated by tritium decay cannot usually escape from solids. Namely, the beta particle of tritium decay is difficult to detect in solids. Conversely, the X-rays induced by beta particles in solids escape from the solid, and they can then be detected outside the solid. This is the basis of beta-ray induced X-ray (BIX) spectrometry (BIXS) of tritium. BIXS has potential for quantitative tritium analysis, but it is qualitative rather than quantitative for the solid phase. Because the number and energy of detected X-ray photons depend on the distribution of tritium in the solid, the shape of the BIX spectrum reflects the tritium distribution in the solid. For a quantitative tritium analysis, the tritium distribution in the solid would be evaluated by the shape of the BIX spectrum. In this study, Monte Carlo simulations were employed to determine the change in the BIX spectrum with the tritium depth. The relative intensity of the characteristic X-ray had the information about the tritium distribution and constituent elements of both a solid and an atmosphere. However, the change in the BIX spectrum with the tritium depth depended on the solid material. To consider the BIX spectrum of an unknown solid sample containing tritium, Monte Carlo simulation of the BIX spectrum is indispensable.

Keywords:Beta-ray induced X-ray, Tritium depth, Monte Carlo simulation, Geant4
Accepted:25 June 2021

[2021_16]

Suitability of a simple sampler using a brass bar for gaseous tritiated water measurement

Masato Nakayama a,*, Masanori Hara a, Fumihiko Kobayashi b, Sachiko Oyama b, Masashi Ota b,Takaharu Sakajo b, Hiroo Nakagawa b, Masato Kondo b

a Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama
b Chiyoda Technol., Inc.

Abstract
A simple sampler using a brass bar and liquid nitrogen for measuring tritium in workplaces was designed, and the suitability of the device was evaluated. The brass bar was arranged to allow its end to touch liquid nitrogen in a stainless-steel bottle, and the water vapor in the air became saturated and condensed on the surface of the brass bar. The condensation was defrosted and the tritium in the defrosted water was analyzed by a liquid scintillation counter. The data obtained with the new sampler were compared with the data obtained with a conventional modified oxidation–liquid collection sampler. The relative ratios of the data obtained with the new sampler to the data obtained with the conventional sampler were 0.64–1.54 (average 0.99, relative standard deviation 23.0%), and the new sampler showed high usefulness in working environment measurement of tritium. The data obtained with the new sampler, as the conventional sampler, showed the seasonal effect and the position dependence in the room. Thus, the applicability of the new sampler for working environment measurement of tritium was confirmed.

Keywords:Working environment measurement, Brass bar, Liquid nitrogen, Cooling condensation, Liquid scintillation counter
Accepted:8 June 2021

[2021_17]

Tritium behavior in isotropic graphite at room temperature

Naritoshi Kawamura *,a,b, Shiro Matoba a,b, Shunsuke Makimura c,d, Masanori Hara e, Yuji Hatano e

a Muon Science Section, Materials and Life Science Division, J-PARC Center
b Muon Science Laboratory, Institute of Materials Structure Science, High Energy Accelerator Research Organization (KEK)
c Hadron Section, Particle and Nuclear Physics Division, J-PARC Center
d Institute of Particle and Nuclear Studies, High Energy Accelerator Research Organization (KEK)
e Faculty of Science, Academic Assembly, University of Toyama

Abstract
For a high power proton accelerator like J-PARC (Japan Proton Accelerator Research Complex), radiation safety is a crucial issue for stable and continuous operation. Tritium is generated by a spallation process even in low-Z materials such as Be and C which are generally applied not only to target materials generating secondary particles but also to other devices irradiated by high energy particles. In general, tritium diffuses easily in a material at high temperature, goes out to the surrounding environment and eventually increases the risk of radiation safety. Thus, tritium is an issue in the maintenance work in short term, and also in storing the spent devices, especially the spent targets in long term due to tritium leakage from and accumulation in a storing container. In J-PARC, isotropic graphite is adopted as the target material for muon production, and thus it is one of the major tritium sources. The present work aims to studying tritium behavior in isotropic graphite to acquire the knowledge applicable to the facility operation.

Keywords:Accelerator facility, Muon production target, Graphite, Tritium
Accepted:29 June 2021

[2021_18]

Cracking behavior and microstructural, mechanical and thermal characteristics of tungsten–rhenium binary alloys fabricated by laser powder bed fusion

Takafumi Yamamoto a,b,*, Masanori Hara c, Yuji Hatano c
a Toyama Industrial Technology Research and Development Center
b Graduate School of Science and Engineering, University of Toyama
c Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama

Abstract
In this study, the cracking behavior and microstructural, mechanical and thermal characteristics of tungsten–rhenium (W–Re) binary alloys fabricated by laser powder bed fusion (L-PBF) were investigated. Four bulk specimens were prepared by L-PBF: pure W, W–1%Re, W–3%Re and W–10%Re (percentages indicate the mass percent of Re). High-density bulk specimens (relative density > 98.0%) were obtained for pure W and W–Re alloys under the same laser irradiation conditions. The columnar grains elongated along the building direction were gradually refined as the Re content increased. The most remarkable grain refinement was observed for the W–10%Re alloy. Hardness under a high-temperature environment increased with increasing Re content; the micro-Vickers hardnesses of pure W and W–10%Re at 400 ◦C were 179 ± 4 HV0.1/30 and 281 ± 5 HV0.1/30, respectively. Observations with a scanning electron microscope revealed that the 10 mass% Re addition resulted in a shorter and narrower crack morphology in comparison with pure W and consequently reduced crack area by 59%. Furthermore, the anisotropy of the thermal diffusivity was mitigated in the high Re content specimens, suggesting that, at high Re content, thermal diffusivity is affected less by cracks than by the effect of Re atoms on heat carrier transfer via isotropic scattering.

Keywords:Additive manufacturing, Laser powder bed fusion, Tungsten, Rhenium alloying, Crack suppression
Accepted:26 July 2021

[2021_19]

Imaging and Characterization of Ni@OLC Synthesized by Microwave-Assisted Catalytic Methane Decomposition

Fumihiro Kodera,*[a] Mitsuhiro Inoue,[b] Saito Nobuo,[c] Minoru Umeda,[c] and Akihiko Miyakoshi[a]

[a] Assoc. Prof. Dr. F. Kodera, Prof. Dr. A. Miyakoshi Department of Materials Chemistry, National Institute of Technology, Asahikawa College
[b] Dr. M. Inoue  Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama
[c] Assoc. Prof. Dr. S. Nobuo, Prof. Dr. M. Umeda  Department of Materials Science and Technology, Faculty of Engineering, Nagaoka University of Technology

Abstract
Herein, we propose a new microwave synthesis method based on a solid-gas phase reaction. The internal structure of onionlike carbon (OLC) containing Ni particle synthesized via the microwave-assisted catalytic methane decomposition is characterized. Surface morphologies are tested by Raman spectroscopy and scanning electron miscroscope. Internal structural analysis is conducted using (i) scanning transmission electron microscopy with energy-dispersive X-ray spectroscopy (EDX) and (ii) the cross-section processing of individual particles using a focused ion beam, followed by the analysis of the crosssectioned particles using field-emission scanning electron microscopy with EDX. Consequently, the components of the synthesized particulate powder are classified into three groups:core-shell-structured Ni@OLC, Ni-dispersed carbon particles, and other components. The core-shell-structured Ni@OLC is composed of carbon particles with predominant Ni cores. The Ni-dispersed carbon particles comprised carbon particles with dispersed Ni and Si. These particles are considered precursors of Ni@OLC.

Keywords:Microwave chemistry, Heterogeneous, catalysis, Transition metals, Metal-organic frameworks, Surface analysis
Accepted:September 6, 2021

[2021_20]

Penetration of deuterium into neutron-irradiated tungsten under plasma exposure

Miyuki Yajima1, Yuji Hatano2, Vladimir Kh Alimov3, Takeshi Toyama4, Tatsuya Kuwabara5, Thomas Schwarz-Selinger6, Yasuhisa Oya7, Alexander V Spitsyn8 and Noriyasu Ohno5

1 National Institute for Fusion Science, Oroshi-cho, Toki, 509-5292, Japan
2 Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan
3 A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russia
4 Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313, Japan
5 Graduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan 6 Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, Garching D-85748, Germany 7 Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan 8 National Research Centre 'Kurchatov Institute', Moscow 123182, Russia
Hydrogen isotope trapping at lattice defects in neutron-irradiated tungsten (W), a leading candidate as plasma facing material, is an important problem determining tritium (T) inventory in a vacuum vessel of a future fusion reactor. In this study, W samples were irradiated with neutrons in the Belgium Reactor 2 at 563 K to 0.06 or 0.016 displacement per atom (dpa). After characterizing defects by positron lifetime measurements, deuterium (D) penetration under exposure to D plasma was examined at 563-773 K. Positron lifetime showed the presence of dislocations, monovacancies and relatively large vacancy clusters. These defects trapped D atoms with different values of binding energy. Dependence of D retention on plasma exposure temperature and damage level indicated that the concentrations of weak traps with smaller binding energy increased more significantly with damage level than those of strong traps.

Keywords:
Accepted:

[2021_21]

解説:多孔質材料を用いた触媒の新展開
 多孔体材料による同位体の分離

田口 明

同位体の需要の高まりを受け,分離材料,分離技術の開発が活発になっている.本稿では,主に水素同位体(H2-D2),トリチウム水(HTO),6Li-7Li分離について,ゼオライト,金属有機構造体(MOF),共有結合性有機構造体(COF)などによる同位体分離の発現や分離能の向上,応用例について,最近の研究例を紹介する.

Keywords:Isotope separation, Hydrogen, Tritium, Porous materials
発行年月日:2021年8月10日