Suppression of vacancy formation and hydrogen isotope retention in irradiated tungsten by addition of chromium

Wang, J.*; Hatano, Yuji*; Toyama, Takeshi*; Suzudo, Tomoaki; Hinoki, Tatsuya*; Alimov, V. Kh.*; Schwarz-Selinger, T.*

Journal of Nuclear Materials, 559, p.153449_1 - 153449_7, 2022/02

https://doi.org/10.1016/j.jnucmat.2021.153449

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-0.3Cr alloy were irradiated with 6.4 MeV Fe ions in the temperature range of 523-1273 K. These displacement-damaged samples were exposed to D gas at a temperature of 673 K. The addition of 0.3% 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. Positron lifetime for W-0.3Cr 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 by 0.3% Cr addition.

Effect of neutron energy and fluence on deuterium retention behaviour in neutron irradiated tungsten

Fujita, Hiroe*; Yuyama, Kenta*; Li, X.*; Hatano, Yuji*; Toyama, Takeshi*; Ota, Masayuki; Ochiai, Kentaro; Yoshida, Naoaki*; Chikada, Takumi*; Oya, Yasuhisa*

Physica Scripta, 2016(T167), p.014068_1 - 014068_5, 2016/02

https://doi.org/10.1088/0031-8949/T167/1/014068

The irradiation defects were introduced by Fe irradiation, fission neutron irradiation and D-T neutron irradiation. After the irradiation, the deuterium ions (D) implantation was performed and the D retention behavior was evaluated by thermal desorption spectroscopy. The experimental results indicated that dense vacancies and voids within the shallow region near the surface were introduced by Fe irradiation. The trapping state of D by vacancies and void were clearly controlled by the damage concentration and the voids would become the most stable D trapping site. For fission neutron irradiated W, most of the D was adsorbed on the surface and/or trapped by dislocation loops and no vacancies and voids for D trapping due to its lower damage concentration. D trapping by vacancies were found in the bulk of D-T neutron irradiated W, indicating that the neutron energy distribution could make a large impact on irradiation defect formation and the D retention behavior.

Tritium distribution on the tungsten surface exposed to deuterium plasma and then to tritium gas

Isobe, Kanetsugu; Alimov, V. Kh.*; Taguchi, Akira*; Saito, Makiko; Torikai, Yuji*; Hatano, Yuji*; Yamanishi, Toshihiko

Journal of Plasma and Fusion Research SERIES, Vol.10, p.81 - 84, 2013/02

http://www.jspf.or.jp/JPFRS/PDF/Vol10/jpfrs2013_10-081.pdf

The distribution of hydrogen trapping sites on W surface exposed with D plasma was examined by the techniques of imaging plate and autoradiography. Recrystallized W specimens were exposed with D plasma at around 495 and 550 K to the same fluence of 10 D/m. Then, tritium was introduced into specimen by the exposure to tritium gaseous at 473 K. After that, the tritium distribution on W surface was examined by the techniques of imaging plate and autoradiography. From the results of the imaging plate, tritium was found to be highly concentrated within the area exposed with D plasma and the concentration of tritium was slightly varied even in that area. In the autoradiograph of W surface, it was found that tritium concentrated on the grain boundary and blisters.

Development of beryllium material for reflector lifetime expansion

Dorn, C. K.*; Tsuchiya, Kunihiko; Hatano, Yuji*; Chakrov, P.*; Kodama, Mitsuo*; Kawamura, Hiroshi

Proceedings of 5th International Symposium on Material Testing Reactors (ISMTR-5) (Internet), 9 Pages, 2012/10

http://www.murr.missouri.edu/ismtr/papercall/Papers/JAEA/Tsuchiya_ISMTR-5_instruments(Takeuchi).pdf

The JMTR has used beryllium reflector since it began operation in 1968. Beryllium has been used as the reflector element material in the reactor, specifically S-200F structural grade beryllium in JMTR. As a part of the reactor upgrade, the Japan Atomic Energy Agency (JAEA) has carried out the cooperation experiments to extend the operating lifetime of the beryllium reflector elements. Thus, three kinds of beryllium metals such as S-200F, S-65H and I-220H were selected at the viewpoints of production methods, impurities and grain size of beryllium starting powders, mechanical properties. Now, data of the material properties and interaction between pure water and these beryllium grades are accumulated under un-irradiated. Additionally, irradiation tests have been prepared and development of PIE technologies has been performed. In this paper, the results of various properties and irradiation test plan for lifetime expansion of beryllium are described for material testing reactors.

Tritium absorption of co-deposited carbon films

Nobuta, Yuji*; Yamauchi, Yuji*; Hino, Tomoaki*; Akamaru, Satoshi*; Hatano, Yuji*; Matsuyama, Masao*; Suzuki, Satoshi; Akiba, Masato

Fusion Engineering and Design, 87(7-8), p.1070 - 1073, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.080

Transfer of tritium in concrete coated with hydrophobic paints

Fukada, Satoshi*; Edao, Yuki*; Sato, Koichi*; Takeishi, Toshiharu*; Katayama, Kazunari*; Kobayashi, Kazuhiro; Hayashi, Takumi; Yamanishi, Toshihiko; Hatano, Yuji*; Taguchi, Akira*; et al.

Fusion Engineering and Design, 87(1), p.54 - 60, 2012/01

https://doi.org/10.1016/j.fusengdes.2011.09.001

An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (1) how fast HTO penetrates through concrete walls, (2) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (3) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The epoxy paint coating can work as a HTO diffusion barrier and the PRF value is around 1/10. The silicon paint coating cannot work as the anti-T permeation barrier, because water deteriorates contact between the paint and cement or mortar.

Surface analysis for chemical combination of beryllium metals

Tsuchiya, Kunihiko; Hanawa, Yoshio; Kitagishi, Shigeru; Ito, Masayasu; Hatano, Yuji*; Matsuyama, Masao*

Heisei-21-Nendo Toyama Daigaku Kyodo Riyo, Kyodo Kenkyu Seika Hokokusho, p.9 - 10, 2010/12

no abstracts in English

Hydrogen production in the -radiolysis of aqueous sulfuric acid solutions containing AlO, SiO, TiO or ZrO fine particles

Yamada, Reiji; Nagaishi, Ryuji; Hatano, Yoshihiko; Yoshida, Zenko

International Journal of Hydrogen Energy, 33(3), p.929 - 936, 2008/02

https://doi.org/10.1016/j.ijhydene.2007.11.028

Hydrogen production was studied in the -radiolysis of aqueous HSO solutions containing oxide powder of AlO, SiO, TiO or ZrO. The observed yields of final product H increased with relative amounts of oxide powder in the solutions and exhibited a particular HSO concentration dependence, which was different for each oxide species and its amount. The addition of a small amount of CHOH to a HSO aqueous solution with oxide powder was quite effective for increasing the final product yields of H. The obtained results revealed that heterogeneous systems composed of oxide powder and aqueous HSO solution were more efficient for producing H molecules in -radiolysis than homogeneous systems without oxides.

Radiation-induced catalytic reduction of chromium(VI) in aqueous solution containing TiO, AlO or SiO fine particles

Nagaishi, Ryuji; Yoshida, Zenko; Yamada, Reiji; Hatano, Yoshihiko

Radiation Physics and Chemistry, 75(9), p.1051 - 1054, 2006/09

https://doi.org/10.1016/j.radphyschem.2006.01.005

Reduction of chromium(VI) in aqueous neutral or basic solution was promoted by -ray irradiation in the presence of oxide particles such as TiO, AlO or SiO. The oxide particles behaved as a catalyst, and the efficiency of the Cr(VI) reduction increased with an increase of the irradiation dose irrespective of the initial Cr(VI) concentration. The insoluble Cr(III) oxide formed through the Cr(VI) reduction also acted as the catalyst.

Radiation-induced reduction process of hazardous chromium(VI) ion to non-hazardous chromium oxide with the lower oxidation state in aqueous solution - oxide system

Nagaishi, Ryuji; Yamada, Reiji; Hatano, Yoshihiko; Yoshida, Zenko

no journal, , 

Radiation-induced reduction of chromium(VI) ion in aqueous solution was investigated in the presence of small amount of insoluble oxide particle such as TiO, AlO and SiO by using Co -ray. Solution at pH 1.0, simulating industrial liquid waste, became neutral or basic by adding the oxide particle: the positive shift of pH of solution was dependent on the kind and particle size of oxide. The chromium(VI) ion was successfully reduced in the radiolysis of the neutral or basic solution in the presence of oxide, while not in the absence. These indicate that amount of chromium(VI) reduced by reducing radicals of water radiolysis is almost equal to that of chromium with the lower oxidation state oxidized by the oxidizing radicals in the solution, and that adsorption of chromium to oxide and enhancement in the reduction of chromium(VI) take place in the presence of oxide. The insoluble chromium(III) oxide formed in the reduction of chromium(VI) was also responsible for the reduction. The radiation-induced reduction process of hazardous chromium(VI) with utilizing high-level radioactive wastes as radiation resources was further discussed.