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Journal Articles

R&D activities of tritium technologies on Broader Approach in Phase 2-2

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

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

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.

Journal Articles

Effect of tritium on corrosion behavior of chromium in 0.01N sulfuric acid solution

Oyaizu, Makoto; Isobe, Kanetsugu; Hayashi, Takumi

Fusion Science and Technology, 67(3), p.519 - 522, 2015/04

 Times Cited Count:2 Percentile:22.25(Nuclear Science & Technology)

The effects of tritiated water on the corrosion behavior of chromium were electrochemically studied by anodic polarization measurements with changing tritium concentration and dissolved oxygen concentration as parameters in the electrolyte of 0.01N sulfuric acid solution, self-passivation due to dissolved oxygen could be observed in pure water without tritium. As a result, it was found that the self-passivation was inhibited in tritiated electrolyte as shown in the previous studies for SUS304 stainless steel. It is indicated from the result that the passivation inhibitory effect for SUS304 stainless steel could be induced by dissolution of chromium in passivation film on SUS304 stainless steel.

Journal Articles

Recent progress on tritium technology research and development for a fusion reactor in Japan Atomic Energy Agency

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

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Hydrogen isotope behavior on a water-metal boundary with simultaneous transfer from and to the metal surface

Hayashi, Takumi; Isobe, Kanetsugu; Nakamura, Hirofumi; Kobayashi, Kazuhiro; Oya, Yasuhisa*; Okuno, Kenji*; Oyaizu, Makoto; Edao, Yuki; Yamanishi, Toshihiko

Fusion Engineering and Design, 89(7-8), p.1520 - 1523, 2014/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Tritium confinement is the most important safety issue in the fusion reactor. Tritium behavior on the water metal boundary is very important to design tritium plant with breading blanket system using cooling water. A series of tritium permeation experiment into pressurized water or water vapor jacket with He or Ar have been performed through pure iron piping with/without 7 micro-meter gold plating, which contained about 1 kPa of pure tritium gas at 423 K, with monitoring the chemical forms of tritium. Also, deuterium permeation experiments from heavy water vessel through various metal piping, such as pure iron (Fe), nickel (Ni), stainless steel (SS304), and pure iron with 10 micro-meter gold plating, were performed at 573 K and at 15 MPa. Recently, using the above heavy water system, we have succeeded to detect simultaneous hydrogen isotopes transfer from and to the metal surface by introducing H$$_{2}$$ gas to the metal piping after stabilized deuterium permeation was detected.

Journal Articles

Recent results on tritium technology in JAEA under BA program

Yamanishi, Toshihiko; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu

Fusion Engineering and Design, 88(9-10), p.2272 - 2275, 2013/10

 Times Cited Count:2 Percentile:21.27(Nuclear Science & Technology)

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 $$gamma$$ irradiation.

Journal Articles

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

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$$^{26}$$ D/m$$^{2}$$. 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.

Journal Articles

Effects of tritiated water on passivation behavior of SUS304 stainless steel

Oyaizu, Makoto; Isobe, Kanetsugu; Yamanishi, Toshihiko

ECS Transactions, 50(50), p.63 - 69, 2013/00

 Times Cited Count:3 Percentile:85.38

The effects of tritiated water on the passivation behavior of SUS304 stainless steel were electrochemically studied by anodic polarization measurements and diachronic measurements of open circuit potential with changing tritium concentration and dissolved oxygen concentration as parameters in the electrolyte of 1N sulfuric acid solution, where the passivation inhibitory effects by tritiated water could be clearly observed. As a result, it was found that the passivation would be proceed with two steps. The effects of tritiated water could be observed in both of two steps; delay in the first step and deceleration in the second step. From these results, it was suggested that the passivation inhibitory effect might be promoted by further oxidation and sequential dissolution of Cr$$^{3+}$$ by radiolysis products.

Journal Articles

Overview of R&D activities on tritium processing and handling technology in JAEA

Yamanishi, Toshihiko; Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Oyaizu, Makoto; Yamada, Masayuki; Suzuki, Takumi; Hayashi, Takumi

Fusion Engineering and Design, 87(5-6), p.890 - 895, 2012/08

 Times Cited Count:1 Percentile:11.68(Nuclear Science & Technology)

In JAEA, the tritium processing and handling technologies have been studied at TPL. The main basic R&D activities in this field are: the tritium processing technology for the blanket recovery system; the tritium behavior in a confinement; and detritiation and decontamination. The R&D for tritium processing and handling technologies to a demonstration reactor (DEMO) are also planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities. The ceramic electrolysis cell has been studied as a tritium processing method for the blanket system. The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column.

Journal Articles

Hydrogen isotope permeation from cooling water through various metal piping

Hayashi, Takumi; Nakamura, Hirofumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Oyaizu, Makoto; Yamanishi, Toshihiko; Oya, Yasuhisa*; Okuno, Kenji*

Fusion Engineering and Design, 87(7-8), p.1333 - 1337, 2012/08

 Times Cited Count:7 Percentile:53.66(Nuclear Science & Technology)

In order to investigate the behavior of hydrogen isotope on the water-metal boundary, deuterium permeation experiments from heavy water vessel through various metal piping, such as pure iron (Fe), nickel (Ni), stainless steel (SS304), and pure iron with 10 $$mu$$m gold plating, were performed at 573 K and at 15 MPa. During the experiment, surfaces of metal piping except gold plating one were oxidized at the heavy water boundary and then deuterium would generate by the oxidation reactions. This deuterium could be detected by mass spectrometer, which monitored the inside gases of the piping under vacuum. The result showed clearly that the deuterium permeated through Fe, Ni, and SS304 piping was detected as deuterium gas (D$$_{2}$$) under vacuum, though that through gold plating one could not be detected effectively. The D$$_{2}$$ permeation rate through Fe, Ni, and SS304 piping reached some stabilized value. This paper summarizes the above experimental results and discusses the mechanism of deuterium behavior on the water metal boundary.

Journal Articles

Effects of tritiated water on corrosion behavior of SUS304

Oyaizu, Makoto; Isobe, Kanetsugu; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 60(4), p.1515 - 1518, 2011/11

 Times Cited Count:3 Percentile:28.8(Nuclear Science & Technology)

The effects of tritiated water on the corrosion behavior of SUS304 stainless steel was studied using Tafel extrapolation method, one of electrochemical techniques with changing tritium concentration, dissolved oxygen concentration and pH in electrolyte as parameters. It was indicated that there would be two or more effects of tritium that enhance the corrosion of SUS304 stainless steel under several experimental conditions. One is passivation inhibitory effect, which could be observed only in highly corrosive circumstance of 1N H$$_{2}$$SO$$_{4}$$ electrolyte. The other effects of tritium on corrosion behavior could be observed not only in 1N H$$_{2}$$SO$$_{4}$$ but also in corrosive circumstance of 1N Na$$_{2}$$SO$$_{4}$$ electrolyte, which would be affected by dissolved oxygen concentration as well as tritium concentration.

Journal Articles

Development of high efficiency electrode for highly tritiated water processing

Isobe, Kanetsugu; Yamanishi, Toshihiko

Fusion Science and Technology, 60(4), p.1387 - 1390, 2011/11

 Times Cited Count:1 Percentile:11.72(Nuclear Science & Technology)

Aiming to enhance the efficiency of ceramic electrolysis method, we developed new electrodes using cerium oxide (Ceria). We prepared electrodes by two manufacturing methods. One is to mix ceria into Pt paste and then electrode was sintering on the YSZ. The other is to use Ceria as intermediate layer between YSZ and Pt-YSZ electrode. The water decomposition performance of such electrodes and usual electrode using Pt-YSZ was confirmed in different humidity at 1073 K. Both electrodes using Ceria showed higher water decomposition performance than that of usual electrode. Especially, 30% ceria adding electrode showed highest performance and the decomposition efficiency was one order magnitude higher than that of usual electrode.

Journal Articles

Radiochemical reactions between tritium and carbon dioxide at elevated temperatures

Isobe, Kanetsugu; Nakamura, Hirofumi; Nakamichi, Masaru; Yamanishi, Toshihiko

Fusion Science and Technology, 60(4), p.1584 - 1587, 2011/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

We focused on the reaction between tritium and carbon dioxide at elevated temperature, self-radiation reactions at 373, 473 and 573 K were investigated. Self-radiation experiment using high purity gaseous tritium was carried out in the chamber of stainless steel at atmospheric pressure and initial ratio between gaseous tritium and carbon dioxide was almost 1:1. After 2 weeks experiment, gas contents and these concentrations were measured with quadrupole mass spectrometer. Main products were carbon monoxide, water and methane. And production ratios of such products were almost same at any temperature. Therefore, it was found that self-radiation reaction between gaseous tritium and carbon dioxide is independent of temperature in the rage of 373-573 K.

Journal Articles

Recent activities of R&D on effects of tritium water on confinement materials and tritiated water processing

Yamanishi, Toshihiko; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Hara, Masanori*; Sugiyama, Takahiko*; Okuno, Kenji*

Fusion Engineering and Design, 86(9-11), p.2152 - 2155, 2011/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

It is quite significant subject how to confine the tritium in a fusion reactor. Especially, it is strongly desired to get the data for tritiated water. This is because tritiated water is much hazardous than the hydrogen form of tritium. As for the behavior of high concentration tritium water, we could get a series of valuable data for the corrosion of the tritiated water against metal materials. In the case where a metal material is in water, an oxidized layer is formed at the surface of the metal. The oxidized layer functions as a passive layer for the corrosion. However, it has been observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). The chemical exchange column has been applied in ITER as the tritium recovery system from tritiated water. A set of data for an advanced chemical exchange column has been obtained.

Journal Articles

Past 25 years results for large amount of tritium handling technology in JAEA

Yamanishi, Toshihiko; Yamada, Masayuki; Suzuki, Takumi; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Inomiya, Hiroshi; Hayashi, Takumi

Fusion Science and Technology, 60(3), p.1083 - 1087, 2011/10

 Times Cited Count:2 Percentile:20.43(Nuclear Science & Technology)

Tritium Process Laboratory (TPL) in Japan Atomic Energy Agency has been established as the only test facilities to handle over 1 gram of in Japan. From March 1988, TPL has been operated with tritium, and no tritium release accident has been observed. The average tritium concentration in a stream from a stack of the TPL to environment was 71 Bq/m$$^{3}$$, and was 1/70 of the Japanese regulation value for HTO. The failure data have been analyzed for several main components of the safety systems such as pumps, valves, and monitors. The data on the tritium waste and accountancy has also been accumulated. As a study of the Grants-in-Aid for Scientific Research, these data are analysed and are reported.

Journal Articles

Temperature dependence of surface topography and deuterium retention in tungsten exposed to low-energy, high-flux D plasma

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

 Times Cited Count:67 Percentile:98.62(Materials Science, Multidisciplinary)

Blistering and deuterium retention in re-crystallized tungsten exposed to a low energy (38 eV/D) and high deuterium ion flux (10$$^{22}$$ D/m$$^{2}$$s) D plasma at ion fluences of 10$$^{26}$$ and 10$$^{27}$$D/m$$^{2}$$ 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 7$$times$$10$$^{21}$$ D/m$$^{2}$$ at 530 K and about 1$$times$$10$$^{22}$$ D m$$^{2}$$ at 480 K for ion fluences of 10$$^{26}$$ and 10$$^{27}$$ D/m$$^{2}$$, respectively. As the temperature grows further, the D retention decreases to about 10$$^{19}$$ D/m$$^{2}$$ at 800 K.

Journal Articles

Permeation behavior of tritium through F82H steel

Oyaizu, Makoto; Isobe, Kanetsugu; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko

Journal of Nuclear Materials, 417(1-3), p.1143 - 1146, 2011/10

 Times Cited Count:5 Percentile:42.62(Materials Science, Multidisciplinary)

The tritium permeation behavior through F82H steel from carrier-free tritiated water vapor was investigated and the difference between as-received sample surface and that after the permeation experiments were analyzed by means of SEM, EDX and XRD. As the results, the permeability of tritium through F82H steel from tritiated water vapor is two to three orders of magnitude smaller than that from gaseous hydrogen. A thick and porous iron oxide layer composed of hematite and magnetite was formed in the permeation experiments. These results indicate that the oxide layer would hardly work as tritium permeation barrier, and that gaseous hydrogen could be generated by the redox reaction of water on the sample surface. Therefore, it could be considered that the tritium permeation through F82H steel from tritiated water vapor could result from the partial pressure of T$$_{2}$$.

Journal Articles

Deuterium retention in porous vacuum plasma-sprayed tungsten coating exposed to low-energy, high-flux pure and helium-seeded D plasma

Alimov, V.; Tyburska, B.*; Ogorodnikova, O. V.*; Roth, J.*; Isobe, Kanetsugu; Yamanishi, Toshihiko

Journal of Nuclear Materials, 415(Suppl.1), p.S628 - S631, 2011/08

 Times Cited Count:12 Percentile:71.79(Materials Science, Multidisciplinary)

Deuterium retention in porous vacuum plasma spraying tungsten coating exposed at various temperatures to a low-energy, high flux (10$$^{22}$$ D/m$$^{2}$$s) pure D and helium-seeded D plasmas was examined by thermal desorption spectroscopy and the nuclear reaction. Under exposure to pure D plasma (76 eV D$$_{2}$$$$^{+}$$) at 340-560 K, the D concentration reaches 1$$sim$$2 at.% at depths of several micrometers, while at temperatures above 700 K the D concentration is below 10$$^{-2}$$ at.% and deuterium is retained over the whole thickness of the coating. Seeding of 76 eV He$$^{+}$$ into the D plasma reduces the D retention at temperatures of 400-600 K. However, at temperature above 700 K, the D retention becomes comparable to that for pure D plasma exposure.

Journal Articles

Hydrogen isotope behavior transferring through water metal boundary

Hayashi, Takumi; Nakamura, Hirofumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Oyaizu, Makoto; Oya, Yasuhisa*; Okuno, Kenji*; Yamanishi, Toshihiko

Fusion Science and Technology, 60(1), p.369 - 372, 2011/07

 Times Cited Count:2 Percentile:20.43(Nuclear Science & Technology)

Journal Articles

Effect of tritium and dissolved oxygen on anodic polarization of SUS304 stainless steel in sulfuric acid solution

Oyaizu, Makoto; Isobe, Kanetsugu; Hayashi, Takumi; Yamanishi, Toshihiko

Journal of Nuclear Science and Technology, 48(6), p.880 - 884, 2011/06

 Times Cited Count:5 Percentile:42.62(Nuclear Science & Technology)

Journal Articles

Tritium concentration in tungsten surface exposed to low-energy, high-flux D plasma

Isobe, Kanetsugu; Alimov, V. Kh.*; Yamanishi, Toshihiko; Torikai, Yuji*

Toyama Daigaku Suiso Doitai Kagaku Kenkyu Senta Kenkyu Hokoku, 31, p.49 - 57, 2011/00

The limits on tritium inventory in the vacuum vessel and the need for prevention of impurity ingress into plasma, make plasma-surface interaction on tungsten an important issue. It is well known that plasma exposure on tungsten makes some kinds of blisters on the surface and increases the hydrogen inventory. On the other hands, there is a possibility that plasma exposure would change the characteristic of surface and surface region in tungsten and cause the increase of tritium inventory. Tritium concentration in tungsten exposure by low-energy, high-flux D plasma with was examined with BIXS after thermal exposure of tritium gas. The tritium concentration was measured with BIXS. The tritium concentration in surface and surface region was found to be increased by plasma exposure. And its concentration of tungsten exposed at 495 K was estimated to be twice higher than that of as-received tungsten.

106 (Records 1-20 displayed on this page)