Otani, Masashi*; Fukao, Yoshinori*; Futatsukawa, Kenta*; Kawamura, Naritoshi*; Matoba, Shiro*; Mibe, Tsutomu*; Miyake, Yasuhiro*; Shimomura, Koichiro*; Yamazaki, Takayuki*; Hasegawa, Kazuo; et al.
Journal of Physics; Conference Series, 1350, p.012067_1 - 012067_6, 2019/12
Negative muonium atom (ee, Mu) has unique features stimulating potential interesting for several scientific fields. Since its discovery in late 1980's in vacuum, it has been discussed that the production efficiency would be improved using a low-work function material. C12A7 was a well-known insulator as a constituent of alumina cement, but was recently confirmed to exhibit electric conductivity by electron doping. The C12A7 electride has relatively low-work function (2.9 eV). In this paper, the negative muonium production measurement with several materials including a C12A7 electride film will be presented. Measured production rate of the Mu were 10/s for all the Al, electride, and SUS target. Significant enhancement on electride target was not observed, thus it is presumed that the surface condition should be more carefully treated. There was no material dependence of the Mu averaged energy: it was 0.20.1keV.
Bae, S.*; Choi, H.*; Choi, S.*; Fukao, Yoshinori*; Futatsukawa, Kenta*; Hasegawa, Kazuo; Iijima, Toru*; Iinuma, Hiromi*; Ishida, Katsuhiko*; Kawamura, Naritoshi*; et al.
Physical Review Accelerators and Beams (Internet), 21(5), p.050101_1 - 050101_6, 2018/05
Muons have been accelerated by using a radio-frequency accelerator for the first time. Negative muonium atoms (Mu), which are bound states of positive muons and two electrons, are generated from through the electron capture process in an aluminum degrader. The generated Mu's are initially electrostatically accelerated and injected into a radio-frequency quadrupole linac (RFQ). In the RFQ, the Mu's are accelerated to 89 keV. The accelerated Mu's are identified by momentum measurement and time of flight. This compact muon linac opens the door to various muon accelerator applications including particle physics measurements and the construction of a transmission muon microscope.
Ueno, Yasuhiro*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.
Hyperfine Interactions, 238(1), p.14_1 - 14_6, 2017/11
Strasser, P.*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.
Hyperfine Interactions, 237(1), p.124_1 - 124_9, 2016/12
Ishiyama, Hironobu*; Jeong, S.-C.*; Watanabe, Yutaka*; Hirayama, Yoshikazu*; Imai, Nobuaki*; Jung, H. S.*; Miyatake, Hiroari*; Oyaizu, Mitsuhiro*; Osa, Akihiko; Otokawa, Yoshinori; et al.
Nuclear Instruments and Methods in Physics Research B, 376, p.379 - 381, 2016/06
Isobe, Kanetsugu; Kawamura, Yoshinori; Iwai, Yasunori; Oyaizu, Makoto; Nakamura, Hirofumi; Suzuki, Takumi; Yamada, Masayuki; Edao, Yuki; Kurata, Rie; Hayashi, Takumi; et al.
Fusion Engineering and Design, 98-99, p.1792 - 1795, 2015/10
Activities on Broader Approach (BA) were started in 2007 on the basis of the Agreement between the Government of Japan and the EURATOM. The period of BA activities consist of Phase1 and Phase2 dividing into Phase 2-1 (2010-2011), Phase 2-2 (2012-2013) and Phase 2-3 (2014-2016). Tritium technology was chosen as one of important R&D issues to develop DEMO plant. R&D activities of tritium technology on BA consist of four tasks. Task-1 is to prepare and maintain the tritium handling facility in Rokkasho BA site in Japan. Task 2, 3 and 4 are main R&D activities for tritium and these are focused on: Task-2) Development of tritium accountancy technology, Task-3) Development of basic tritium safety research, Task-4) Tritium durability test. R&D activities of tritium technology in Phase 2-2 were underway successfully and closed in 2013.
Hoshino, Tsuyoshi; Ochiai, Kentaro; Edao, Yuki; Kawamura, Yoshinori
Fusion Science and Technology, 67(2), p.386 - 389, 2015/03
Demonstration power reactors (DEMOs) require advanced tritium breeders that have high stability at high temperatures. Therefore, the pebble fabrication of LiTiO with excess Li (LiTiO) as an advanced tritium breeder was carried out. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed. DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. The LiTiO pebbles exhibited good tritium release properties similar to the LiTiO pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water.
Hayashi, Takumi; Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Yamada, Masayuki; Suzuki, Takumi; Kurata, Rie; Oyaizu, Makoto; Edao, Yuki; et al.
Fusion Science and Technology, 67(2), p.365 - 370, 2015/03
Edao, Yuki; Kawamura, Yoshinori; Kurata, Rie; Fukada, Satoshi*; Takeishi, Toshiharu*; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Science and Technology, 67(2), p.320 - 323, 2015/03
The present study aims at obtaining fundamental knowledge for tritium transfer behavior and interaction between tritium and paint coated on concrete walls. The amounts of tritium penetration and release in cement paste with epoxy and urethane paint coatings were measured. The tritium penetration amounts were increased with the HTO exposure time. Time to achieve each saturate tritium value was more than 60 days for cement paste coated with epoxy paint and with urethane paint, while cement paste without paint took 2 days to achieve it. Tritium penetration rates were estimated by an analysis of diffusion model. Although their paint coatings were effective for reduction of tritium penetration through the cement paste exposed to HTO for a short period, the amount of tritium trapped in the paints became large for a long time. This work has been performed under the collaboration research between JAEA and Kyushu University.
Fukada, Satoshi*; Katayama, Kazunari*; Takeishi, Toshiharu*; Edao, Yuki; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Science and Technology, 67(2), p.99 - 102, 2015/03
Ishiyama, Hironobu*; Jeong, S.-C.*; Watanabe, Yutaka*; Hirayama, Yoshikazu*; Imai, Nobuaki*; Miyatake, Hiroari*; Oyaizu, Mitsuhiro*; Katayama, Ichiro*; Osa, Akihiko; Otokawa, Yoshinori; et al.
Japanese Journal of Applied Physics, 53(11), p.110303_1 - 110303_4, 2014/11
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.
Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10
The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.
Edao, Yuki; Kawamura, Yoshinori; Yamanishi, Toshihiko; Fukada, Satoshi*
Fusion Engineering and Design, 89(9-10), p.2062 - 2065, 2014/10
Tritium transfer behavior through hydrophobic paints, epoxy resin and acrylic-silicon resin, was investigated experimentally. The authors measured the amount of tritium permeated through the paint membranes which exposed in HTO atmosphere of 2100 Bq/cm. The most of tritium permeated through the paints in the form of HTO at room temperature. Tritium permeation through the acrylic-silicon paint was explained a linear sorption/release model and that through the epoxy paint was suggested to be controlled by a one-dimensional diffusion model. While effective diffusivity was 1.0101.810 m/s at 21C26C for epoxy membrane, the diffusivity was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, tritium diffusivity through interface between cement-paste and the epoxy paint was considered to be most effective in the overall tritium transfer process. Tritium transfer behavior in the interface is important to explain the mechanism of tritium transfer behavior in concrete walls.
Ochiai, Kentaro; Kawamura, Yoshinori; Hoshino, Tsuyoshi; Edao, Yuki; Takakura, Kosuke; Ota, Masayuki; Sato, Satoshi; Konno, Chikara
Fusion Engineering and Design, 89(7-8), p.1464 - 1468, 2014/10
We have performed the tritium recovery experiment on fusion reactor blanket with DT neutrons at the Fusion Neutronics Source facility in Japan Atomic Energy Agency. The candidate breeding material, LiTiO pebble, was put into the container which was set up it into an assembly simulating water cooled ceramic breeding (WCCB) blanket. Helium sweep gas including H (1%) and/or HO (1%) was flowed and extracted tritium was collected to water bubblers during DT neutron irradiation. The LiTiO pebble was also heated up to a constant temperature at 573, 873 and 1073 K, respectively. We arranged the tritium recovery system to measure tritiated water moisture and tritium gas, separately, and to investigate the amount of recovered tritium and the chemical form. From our experiments, it was showed that the amount of recovered tritium was corresponded to the calculation value and the ratio of chemical form depended to the temperature and kinds of sweep gas.
Kawamura, Yoshinori; Edao, Yuki; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Engineering and Design, 89(7-8), p.1539 - 1543, 2014/10
Tritium recovery system using adsorption or catalytic isotope exchange has already been proposed for a solid breeding blanket system of a nuclear fusion reactor. Synthetic zeolite is often used as an adsorbent or a substrate of chemical exchange catalyst. And, it is well known that its properties are changed easily by exchanging their cations. So, in this work, adsorption capacities of hydrogen isotope and water vapor on cation-exchanged mordenite with transition metal ion were investigated. Ag ion-exchanged mordenite (Ag-MOR) has indicated considerably large hydrogen adsorption capacity in lower pressure range at 77 K. And, adsorption capacity of water vapor did not so vary with exchaned cation in comparison with hydrogen adsorption. The discussion from the viewpoint of adsorption rate is still remaining, but more compact cryosorption column for tritium recovery system is possible to design if Ag-MOR is adopted.
Kawamura, Yoshinori; Iwai, Yasunori; Munakata, Kenzo*; Yamanishi, Toshihiko
Journal of Nuclear Materials, 442(1-3), p.S455 - S460, 2013/11
Zeolite easily exchanges its cation to another. In this work, synthetic mordenite type zeolite (Na-MOR) was used as start material. And, its cation (Na) has been exchanged by Li, K, Mg and Ca. Then, adsorption capacities of H and D on them were investigated at 77 K, 159 K, 175 K and 195 K. Adsorption capacities on Li-MOR and Ca-MOR became larger than that on Na-MOR at low pressure range. Oppositely, that on K-MOR became smaller. In case of alkaline metal, cation with small atomic number may lead to large adsorption capacity.
Iwai, Yasunori; Sato, Katsumi; Kawamura, Yoshinori; Yamanishi, Toshihiko
Fusion Engineering and Design, 88(9-10), p.2319 - 2322, 2013/10
The Nafion ion exchange membrane is a key material for electrolysis cells of the water detritiation system. Endurance of ion exchange membrane immersed into high-concentration tritiated water has been demonstrated under the Broader Approach activities, as a R&D on endurance of fuel cycle components at high tritium exposure. Long-term exposure of Nafion ion exchange membrane into 1.38 TBq/kg of tritiated water was conducted at room temperature for up to 2 years. The curves of percent elongation at break vs. dose and tensile strength vs. dose for the Nafion membranes immersed into tritiated water were well consistent with those for Nafion membranes irradiated to an equivalent dose with rays and electron beams. The results of ferric Fenton test indicated that the degradation directly by radiation was dominant at room temperature compared with that by reactions with radicals produced from water radiolysis. The curve of ion exchange capacity vs. dose for the Nafion membranes immersed into tritiated water was also well consistent with that for Nafion membranes irradiated to an equivalent dose with rays and electron beams. These results showed that the irradiation tests with rays and electron beams were effective to predict a degradation behavior of ion exchange membrane immersed into high-concentration tritiated water.
Yamanishi, Toshihiko; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu
Fusion Engineering and Design, 88(9-10), p.2272 - 2275, 2013/10
The multi-purpose RI equipment has been constructed at Rokkasho site in DEMO R&D building until 2011. The equipment is the first and unique facility in Japan, where tritium, RI species, and beryllium can simultaneously be used. The amounts of tritium used and stored are 3.7 TBq per day and 7.4 TBq, respectively. The material of the column of the micro gas chromatograph has been studied. The calorimeter has also been studied as a possible tritium measurement method. A set of basic data on the interaction between materials and tritium has been measured especially for pure Fe. As for the tritium behavior in the blanket materials, the tritium release after neutron irradiation was studied. As a study for the tritium durability, the endurance of ion exchange membrane has been tested by using high concentration tritium water. The data of tritium water were well consistent with those obtained by irradiation.
Kawamura, Yoshinori; Edao, Yuki; Yamanishi, Toshihiko
Fusion Engineering and Design, 88(9-10), p.2255 - 2258, 2013/10
To develop an adsorbent that is suitable for a separation column of gas chromatograph for hydrogen isotope analysis, the mordenite-type zeolite of which cations (Na) were exchanged with other cations have been prepared and their hydrogen isotope adsorption behavior is being investigated. Then, it has been shown experimentally that mordenite-type zeolite of which cation has been exchanged with Ca (Ca-MOR) has fairly large adsorption capacity. So, breakthrough curves of H (or D) adsorption on Ca-MOR at 194 K and 175 K have been observed and mass transfer coefficients have been estimated from them. The rate-controlling step of hydrogen adsorption is hydrogen diffusion in porous adsorbent. And, isotopic difference of effective diffusivity in Ca-MOR is larger than that in Na-MOR. Therefore, in comparison with Na-MOR, use of Ca-MOR is expected to enhance the hydrogen isotope separation capability.
Edao, Yuki; Kawamura, Yoshinori; Ochiai, Kentaro; Hoshino, Tsuyoshi; Takakura, Kosuke; Ota, Masayuki; Iwai, Yasunori; Yamanishi, Toshihiko; Konno, Chikara
JAEA-Research 2012-040, 15 Pages, 2013/02
Tritium generation and recovery studies on LiTiO as a solid breeding material under neutron irradiation carried out in the Fusion Neutron Source (FNS) facility. A capsule with LiTiO packed bed was put in a system which simulated an actual blanket system which built in beryllium blocks and lithium titanate ones. Estimated values of the amount of tritium generation by a numerical calculation agreed closely with experimental values. The capsule was heated up to 300C, and helium, helium with water vapor, hydrogen or hydrogen/water vapor were selected as purge gas. In the case of purge by helium added water vapor, the ratio of HTO to total tritium release was 98%. In helium with hydrogen/water vapor purge, the ratio of HTO to total tritium release was 80%, which was confirmed that HTO released by isotope exchange reaction between water vapor and tritium. In helium with hydrogen purge, the ratio of HT to total tritium release was 6070%, which was shown that HT released by isotope exchange reaction between hydrogen gas and tritium. HTO released by water generation reaction between hydrogen in purge gas and oxygen in LiTiO although water vapor was not added in purge gas. The ratio of HTO release seemed to be small under the deoxidized condition of the LiTiO surface. Tritium release behavior in the LiTiO depended on the composition of purge gas, and its chemical form was affected by the surface conditions of LiTiO.