Kawamura, Yoshinori; Shu, Wataru*; Matsuyama, Masao*; Yamanishi, Toshihiko
Fusion Science and Technology, 60(3), p.986 - 989, 2011/10
Assuming the blanket sweep gas at the outlet of the blanket, tritium gas monitor by -ray induced X-ray spectroscopy has been modified, and has measured tritium at 120 C. The counting rate at 120 C was about 1/2 of that at the room temperature. In this work, the measurement system was a closed system. When two systems have same volume and same pressure, the number of molecules in higher temperature system is smaller. This is one of the causes of small counting rate. The deterioration of the scintillator after heating was not observed.
Alimov, V.; Shu, Wataru*; Roth, J.*; Lindig, S.*; Balden, M.*; Isobe, Kanetsugu; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.572 - 575, 2011/10
Blistering and deuterium retention in re-crystallized tungsten exposed to a low energy (38 eV/D) and high deuterium ion flux (10 D/ms) D plasma at ion fluences of 10 and 10D/m at temperatures in the range from 320 to 800 K have been examined with scanning electron microscopy, thermal desorption spectroscopy (TDS), and the nuclear reaction. During exposure to the D plasma blisters with various shapes and sizes depending on the exposure temperature are formed on the W surface. At the temperatures above 700 K the blisters disappear. The deuterium retention increases with the exposure temperature, reaching its maximum value of about 710 D/m at 530 K and about 110 D m at 480 K for ion fluences of 10 and 10 D/m, respectively. As the temperature grows further, the D retention decreases to about 10 D/m at 800 K.
Shu, L.*; Higemoto, Wataru; Aoki, Yuji*; Hillier, A. D.*; Oishi, Kazuki*; Ishida, Kenji*; Kadono, Ryosuke*; Koda, Akihiro*; Bernal, O. O.*; MacLaughlin, D. E.*; et al.
Physical Review B, 83(10), p.100504_1 - 100504_4, 2011/03
Zero-field muon spin relaxation experiments have been carried out in the Pr(OsRu)Sb and PrLaOsSb alloy systems to investigate broken time-reversal symmetry (TRS) in the superconducting state, signaled by the onset of a spontaneous static local magnetic field B. In both alloy series B initially decreases linearly with solute concentration. Ru doping is considerably more efficient than La doping, with a 50% faster initial decrease. The data suggest that broken TRS is suppressed for Ru concentration 0.6 but persists for essentially all La concentrations. Our data support a crystal-field excitonic cooper pairing mechanism for TRS-breaking superconductivity.
Alimov, V.; Roth, J.*; Shu, Wataru*; Komarov, D. A.*; Isobe, Kanetsugu; Yamanishi, Toshihiko
Journal of Nuclear Materials, 399(2-3), p.225 - 230, 2010/04
A study of the influence of the deposition conditions on the surface morphology and deuterium concentration in tungsten deposition layers formed by magnetron sputtering and in the linear plasma generator has been carried out. Adhesion of the W layer to substrate is shown to depend on the coefficients of thermal expansion for tungsten and substrate material, thickness of the W layer, and the substrate temperature during layer deposition. A decreased D concentration for increased substrate temperatures and deposition rate are observed.
Alimov, V.; Shu, Wataru*; Roth, J.*; Sugiyama, Kazuyoshi*; Lindig, S.*; Balden, M.*; Isobe, Kanetsugu; Yamanishi, Toshihiko
Physica Scripta, T138, p.014048_1 - 014048_5, 2009/12
Blistering and deuterium retention in re-crystallized tungsten exposed to low-energy, high flux (10 D/ms) pure and helium-seeded D plasmas to a fluence of 10 D/m have been examined with scanning electron microscopy, thermal desorption spectroscopy (TDS), and the D(He,p)He nuclear reaction at a He energy varied from 0.69 to 4.0 MeV. In the case of exposure to pure D plasma (38 eV/D), blisters with various shapes and sizes depending on the exposure temperature are found on the W surface. No blisters appear at temperatures above 700 K. The deuterium retention increases with the exposure temperature, reaching a maximum value of about 10 D/m at 480 K, and then decreases as the temperature rises further. Seeding of helium into the D plasma to the He ion concentration of 0.2 and 5% significantly reduces the D retention at elevated temperatures and prevents formation of the blisters.
Shu, Wataru; Nakamichi, Masaru; Alimov, V.; Luo, G.-N.*; Isobe, Kanetsugu; Yamanishi, Toshihiko
Journal of Nuclear Materials, 390-391, p.1017 - 1021, 2009/06
no abstracts in English
Shu, Wataru; Kawasuso, Atsuo; Yamanishi, Toshihiko
Journal of Nuclear Materials, 386-388, p.356 - 359, 2009/04
Tungsten is a most promising plasma facing material because of its high melting point. The blistering and deuterium retention in recrystallized tungsten samples were investigated by using simulated edge-plasma of fusion reactors. High-dome blisters appeared at the tungsten surface after deuterium plasma exposure, and their ratios of height against chord of the blisters were even one-order greater that reported before. In addition, there was a cavity in the inside of small blisters, whereas there was a void/crack along the grain boundary beneath the big blister and there is no lid for big blisters. Besides the strong dependence upon the exposure temperature, blistering and deuterium retention also showed significant dependence upon the features of microstructure.
Sato, Hideyuki*; Aoki, Yuji*; Kikuchi, Daisuke*; Sugawara, Hitoshi*; Higemoto, Wataru; Oishi, Kazuki; Ito, Takashi; Heffner, R. H.; Saha, S. R.*; Koda, Akihiro*; et al.
Physica B; Condensed Matter, 404(5-7), p.749 - 753, 2009/04
Wide varieties of strongly correlated electron phenomena are performed on the stage of a filled skutterudite structure. Especially when one of the players contains a plural number of 4f electrons, the orbital degrees of freedom play a major role as a new type of nonmagnetic and/or weak-magnetic phenomena. Several examples found in Pr- and Sm-based filled skutterudites are introduced in relation to muon spin relaxation experiments.
Alimov, V.; Shu, Wataru; Roth, J.*; Komarov, D. A.*; Lindig, S.*; Isobe, Kanetsugu; Nakamura, Hirofumi; Yamanishi, Toshihiko
Advanced Materials Research, 59, p.42 - 45, 2009/00
Shu, Wataru; Isobe, Kanetsugu; Yamanishi, Toshihiko
Fusion Engineering and Design, 83(7-9), p.1044 - 1048, 2008/12
At 315 K, only sparse low-dome blisters appeared even the fluence was increased to 10 D/m. At around 400 K, the blisters became much denser and the dome of blisters became a little higher. Peculiar change occurred around 500 K, where two kinds of blisters appeared. One is the large blisters with sizes of a few tens of microns and varying ratios of height against chord (up to 0.6), and the other is the small blisters with chords of less than a few microns and large ratio of height against chord (about 0.7). In high temperature region (higher than 600 K), the blisters became much sparser with the increasing temperature and disappeared at 1000 K. Deuterium retention showed the maximum around 500 K, corresponding to the appearance of two kinds of high-dome blisters.
Hayashi, Takumi; Suzuki, Takumi; Yamada, Masayuki; Shu, Wataru; Yamanishi, Toshihiko
Fusion Engineering and Design, 83(10-12), p.1429 - 1432, 2008/12
In ITER facility, about 3 kg of tritium will be stored in more than 30 ZrCo hydride beds, as a reference design. The safe design and operation of tritium storage beds will be one of the most important points to enhance total safety of the facility. In the Tritium Process Lab. in Japan Atomic Energy Agency, many tritium storage beds with ZrCo have been used with/without self-accountancy measure, and the safe handling experiences have been accumulated for almost 20 years. From these experiences, the key issues to be considered for the safety design are the effect of tritium decay, such as decay heat transfer and He behavior with the normal protection of over temperature, over pressure and leak for a metal-hydride bed. Concerning the safety operation, the key issues are the procedure of hydrogenation-dehydrogenation cycle under the requirements of the storage system and the emergency performances, such as a rapid hydrogen recovery and loss of normal cooling function.
Yamanishi, Toshihiko; Hayashi, Takumi; Shu, Wataru; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Arita, Tadaaki; Hoshi, Shuichi; et al.
Fusion Engineering and Design, 83(10-12), p.1359 - 1363, 2008/12
At TPL (Tritium Process Laboratory) of JAEA, ITER relevant tritium technologies have been studied. The design studies of Air Detritiation System have been carried out in JAEA as a contribution of Japan to ITER. For the tritium processing technologies, our efforts have been focused on the research of the tritium recovery system of ITER test blanket system. A ceramic proton conductor has been studied as an advanced blanket system. A series of fundamental studies on tritium safety technologies not only for ITER but also for fusion DEMO plants has also been carried out at TPL of JAEA. The main research activities in this field are the tritium behavior in a confinement and its barrier materials; monitoring; accountancy; detritiation and decontamination etc. In this paper, the results of above recent activities at TPL of JAEA are summarized from viewpoint of ITER relevant and future fusion DEMO reactors.
Yamanishi, Toshihiko; Hayashi, Takumi; Shu, Wataru; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu
Fusion Science and Technology, 54(1), p.45 - 50, 2008/07
The R&D for tritium technologies towards to the DEMO plants are carried out in Broader Approach (BA) program in Japan: (1) tritium accountancy technology; (2) basic tritium safety research; and (3) tritium durability test. A multi-purpose facility is constructed at Rokkasho in Japan to carry out the above R&Ds. Beta radioisotopes as well as tritium (370 TBq/year) can be handled in the facility. At TPL (Tritium Process Laboratory) of JAEA, a series of R&Ds for the tritium technologies relevant to the above BA program have been started. A series of basic studies for the tritium-materials has also been carried out. The main R&D activities in this field are the tritium behavior in a confinement; monitoring; detritiation; and decontamination. In this paper, the results of above recent activities at TPL of JAEA are also summarized from viewpoint of future fusion DEMO reactors.
Yamanishi, Toshihiko; Yamada, Masayuki; Suzuki, Takumi; Shu, Wataru; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Hoshi, Shuichi; et al.
Fusion Science and Technology, 54(1), p.315 - 318, 2008/07
The construction of the building and safety systems of the TPL was completed until 1985. The operations of the safety systems with tritium have been started from March 1988. The amount of tritium held at the TPL was 13 PBq at March 2007. The average tritium concentration in a stream from a stack of the TPL to environment was 6.010 Bq/cm; and is 1/100 smaller than that of the regulation value for the concentration of HTO in the air in Japan. The safety operation results with tritium have thus been obtained. A set of failure data of several main components of the TPL was also obtained as the valuable data for fusion tritium facilities.
Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Isobe, Kanetsugu; Nakamura, Hirofumi; Kawamura, Yoshinori; Shu, Wataru; Suzuki, Takumi; Yamada, Masayuki; Yamanishi, Toshihiko
Fusion Science and Technology, 54(1), p.319 - 322, 2008/07
Applied Physics Letters, 92(21), p.211904_1 - 211904_3, 2008/05
The high-dome blisters appearing on tungsten after deuterium plasma exposure at around 500K are considered to be generated by the diffusion and agglomeration of the deuterium-vacancy clusters, and this kind of deformation is called as deuterium-induced local superplasticity. There were cavities inside the blisters smaller than a few micrometers, whereas there were no hollow lids formed for some blisters greater than a few micrometers, which is contrary to the typical feature of blisters reported before.
Kawamura, Yoshinori; Arita, Tadaaki; Isobe, Kanetsugu; Shu, Wataru; Yamanishi, Toshihiko
Fusion Engineering and Design, 83(4), p.625 - 633, 2008/05
Electrochemical hydrogen pump with ceramic proton conductor membrane has been proposed to apply for a blanket tritium recovery system (BTR) of a fusion reactor. SrCeYbO (SCO) is one of the candidates of membrane. Modification of electrode is one of the methods to enhance the hydrogen transportation capability. In this work, the electrodes of platinum (Pt) and palladium (Pd) were attached to the SCO sample by the sputtering method (sputtering electrode), and its electric conductivity and proton conductivity were measured. Then, they were compared with that of the usual Pt paste electrode. Hydrogen transportation capability was enhanced when the sputtering electrode was applied. Especially, in case of the Pd sputtering electrode, the current density which was about 4 or 5times larger than the usual Pt paste electrode was observed at 0.1% of H concentration.
Kobayashi, Kazuhiro; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Nakamura, Hirofumi; Kawamura, Yoshinori; Shu, Wataru; Suzuki, Takumi; Yamada, Masayuki; Yamanishi, Toshihiko
Proceedings of 2nd Japan-China Workshop on Blanket and Tritium Technology, p.74 - 78, 2008/05
In order to accumulate the tritium behavior in the future fusion reactor included ITER, intentional tritium release experiments have been carried out using Caisson Assembly for Tritium Safety study (CATS) at Tritium Process Laboratory (TPL) in Japan Atomic Energy Agency (JAEA). Main objectives of CATS are (1) to demonstrate the initial tritium behavior in the room and to develop 3D simulation code of tritium behavior in the room. (2) to demonstrate the performance of integrated system for tritium confinement after intentional tritium release accident, (3) to accumulate the data for the detritiation behavior and the interaction between various materials and tritium (tritiated water) in the confinement. The study using CATS has been continued for about 10 yeas in TPL/JAEA.
Oishi, Kazuki; Heffner, R. H.; Ito, Takashi; Higemoto, Wataru; Morris, G. D.*; Bauer, E. D.*; Graf, M. J.*; Zhu, J.-X.*; Morales, L. A.*; Sarrao, J. L.*; et al.
Physica B; Condensed Matter, 403(5-9), p.1013 - 1014, 2008/04
PuCoGa has attracted much interest because it is the first Pu-based superconductor, having an order of magnitude higher transition temperature K than the isostructural heavy fermion superconductor CeCoIn (K). The mechanism of the superconductivity in PuCoGa is still under investigation, though recent experiments and theory suggest a magnetic origin. A unique aspect of this compound is the self-irradiation damage because Pu (Pu, = 24,000 years) creates lattice defects which scatter electrons and, hence, break superconducting pairs. In order to elucidate the magnitude and temperature dependence of the magnetic penetration depth , we have performed SR measurements in the same PuCoGa single crystals after 25 and 400 days of aging. We found that decreased from 18.5K to 15K for the aged sample, yet a quasi-linear temperature dependence was found for the low-temperature in both the fresh and aged sample, consistent with -wave pairing symmetry. The magnitude of the muon spin relaxation rate in the aged sample, , where and are the superfluid density and the effective mass, respectively, is reduced by about 70% compared to fresh sample. This indicates that the scattering from self-irradiation induced defects is not in the limit of the conventional Abrikosov-Gor'kov pair-breaking theory, but rather in the limit of short coherence length (about 2nm in PuCoGa) superconductivity.
Kobayashi, Kazuhiro; Isobe, Kanetsugu; Iwai, Yasunori; Hayashi, Takumi; Shu, Wataru; Nakamura, Hirofumi; Kawamura, Yoshinori; Yamada, Masayuki; Suzuki, Takumi; Miura, Hidenori*; et al.
Nuclear Fusion, 47(12), p.1645 - 1651, 2007/12
The confinement and removal of tritium are the key subjects for safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell building, tritium is often released, as vapor and is in contact with the inner walls. Also those of an ITER tritium plant building will be exposed to tritium in an accident. However, the data are scarce, especially on the penetration of tritium into the concrete of the wall materials. The tritium released in the buildings is removed by the Atmosphere Detritiation Systems (ADS), where the tritium is oxidized by catalysts and is removed as water. Special gas of SF is used in ITER, and is expected to be released in an accident such as fire. Although the SF gas has the potential as a catalyst poison, the performance of ADS with the existence of SF has not been confirmed yet. Tritiated water is produced in the regeneration process of ADS, and is subsequently processed by the ITER Water Detritiation System (WDS). One of the key components of WDS is an electrolysis cell. The electrolysis cell is made of organic compounds, and there is no data on the durability of the cell exposed to tritium. To overcome these issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R&D task: (1) tritium behavior in concrete; (2) effect of SF on performance of ADS; and (3) tritium durability of electrolysis cell of ITER-WDS.