Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 30

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Additional information to report on site tour of the Fukushima Daiichi Nuclear Power Station

Suto, Toshiyuki

Genshiryoku, hoshasen Bukaiho (Internet), (19), P. 15, 2016/12

The Tritiated Water Task Force under METI's Committee on Countermeasures for Contaminated Water Treatment for Fukushima Daiichi Nuclear Power Plant (1F) reported that the option of post-dilution offshore release could dispose the tritiated water at a smallest cost in the shortest amount of time. The amount of tritium in the contaminated water at 1F was compared with ones released from nuclear power plants and reprocessing plants as some help for grasping its level of magnitude.

Journal Articles

Report on site tour of the Fukushima Nuclear Power Station

Suto, Toshiyuki

Gijutsushi, 28(11), p.8 - 11, 2016/11

Five years have passed since the accident of the Fukushima Daiichi Nuclear Power Station. The Nuclear and Radiation section of the Institute of Professional Engineers hosted a site tour of the plant to make themselves sure what is going on in it and to disseminate information about it. The conditions of landscape during traveling between the gathering place and the plant, each reactor, contaminated water treatment, site, and work environment improvement will be reported.

Journal Articles

Catalyst technology of Tanaka Kikinzoku Kogyo

Kubo, Hitoshi*; Oshima, Yusuke*; Iwai, Yasunori

JETI, 63(10), p.33 - 36, 2015/09

Tanaka Kikinzoku Kogyo provides a broad range of precious metals products and technologies. Tanaka Kikinzoku Kogyo and Japan Atomic Energy Agency have jointly developed a new method of manufacturing catalysts involving hydrophobic processing with an inorganic substance base. As a result, previous technological issues were able to be solved with the development of a catalyst that exhibited no performance degradation in response to radiation application of 530 kGy, a standard for radiation resistance, and maintenance of thermal stability at over 600$$^{circ}$$C, which is much higher than the 70$$^{circ}$$C temperature that is normally used. The application of this catalyst to the liquid phase catalytic exchange process is expected to overcome significant technological hurdles with regards to improving the reliability and efficiency of systems for collecting tritium from tritiated water. It is also anticipated that the hydrophobic platinum catalyst manufacturing technology used for this catalyst could be applied to a wide range of fields other than nuclear fusion research. It was verified that if applied to a hydro oxidation catalyst, hydrogen could be efficiently oxidized, even at room temperature. This catalyst can also contribute to improving safety at non-nuclear plants that use hydrogen in general by solving the aforementioned vulnerability issue.

Journal Articles

Development of hydrophobic platinum catalyst for the effective collection of tritium in fusion plants

Iwai, Yasunori; Kubo, Hitoshi*; Oshima, Yusuke*

Isotope News, (736), p.12 - 17, 2015/08

We have successfully developed a new hydrophobic platinum catalyst for collecting tritium at nuclear fusion reactors. Catalysts used to collect tritium are called hydrophobic precious metal catalysts. In Japan, hydrophobic precious metal catalysts manufactured from polymers have been used for heavy water refinement.However, this catalyst has issues related to embrittlement to radiation and thermal stability. These technological issues needed to be solved to allow for its application to nuclear fusion reactors requiring further enrichment from highly-concentrated tritiated water. We developed a new method of manufacturing catalysts involving hydrophobic processing with an inorganic substance base. As a result, previous technological issues were able to be solved with the development of a catalyst that exhibited no performance degradation in response to radiation application of 530kGy, a standard for radiation resistance, and maintenance of thermal stability at over 600$$^{circ}$$C, which is much higher than the 70$$^{circ}$$C temperature that is normally used. The catalyst created with this method was also confirmed to have achieved the world's highest exchange efficiency, equivalent to 1.3 times the previously most powerful efficiency. The application of this catalyst to the liquid phase catalytic exchange process is expected to overcome significant technological hurdles with regards to improving the reliability and efficiency of systems for collecting tritium from tritiated water.

Journal Articles

Successful development of a new catalyst for efficiently collecting tritium; A Breakthrough toward realization of fusion reactors

Iwai, Yasunori; Kubo, Hitoshi*; Oshima, Yusuke*

Kagaku, 70(5), p.35 - 40, 2015/05

We have successfully developed a new hydrophobic platinum catalyst for collecting tritium at nuclear fusion reactors. Catalysts used to collect tritium are called hydrophobic precious metal catalysts. In Japan, hydrophobic precious metal catalysts manufactured from polymers have been used for heavy water refinement. However, this catalyst has issues related to embrittlement to radiation and thermal stability. These technological issues needed to be solved to allow for its application to nuclear fusion reactors requiring further enrichment from highly-concentrated tritiated water. We developed a new method of manufacturing catalysts involving hydrophobic processing with an inorganic substance base. As a result, previous technological issues were able to be solved with the development of a catalyst that exhibited no performance degradation in response to radiation application of 530 kGy, a standard for radiation resistance, and maintenance of thermal stability at over 600$$^{circ}$$C, which is much higher than the 70$$^{circ}$$C temperature that is normally used. The catalyst created with this method was also confirmed to have achieved the world's highest exchange efficiency, equivalent to 1.3 times the previously most powerful efficiency. The application of this catalyst to the liquid phase catalytic exchange process is expected to overcome significant technological hurdles with regards to improving the reliability and efficiency of systems for collecting tritium from tritiated water.

Journal Articles

Distinctive radiation durability of an ion exchange membrane in the SPE water electrolyzer for the ITER water detritiation system

Iwai, Yasunori; Yamanishi, Toshihiko; Isobe, Kanetsugu; Nishi, Masataka; Yagi, Toshiaki; Tamada, Masao

Fusion Engineering and Design, 81(1-7), p.815 - 820, 2006/02

 Times Cited Count:14 Percentile:24.8(Nuclear Science & Technology)

Solid-polymer-electrolyte (SPE) water electrolysis is attractive in electrolytic process of water detritiation system (WDS) in fusion reactors because it can electrolyze liquid waste directly, but radioactive durability of its ion exchange membrane is a key point. Radioactive durability of Nafion, a typical commercial ion exchange membrane, was experimentally investigated using Co-60 irradiation facility and electron beam irradiation facility at Takasaki Radiation Chemistry Research Establishment of JAERI. Nafion is composed of PTFE (Polytetrafluoroethylene) main chain. However the degradation of its mechanical strength by irradiation was significantly distinguished from that of PTFE and no serious damage was observed for its ion exchange capacity up to 530 kGy, the requirement of ITER. Atmospheric effects such as soaking and oxygen on degrading behaviors were discussed from the viewpoint of radical reaction mechanism. Dependencies of operating temperature and radioactive source are also demonstrated in detail.

Journal Articles

Sorption and desorption of tritiated water on four kinds of materials for ITER

Kobayashi, Kazuhiro; Hayashi, Takumi; Nishi, Masataka; Oya, Yasuhisa*; Okuno, Kenji*

Fusion Engineering and Design, 81(8-14), p.1379 - 1384, 2006/02

 Times Cited Count:4 Percentile:57.94(Nuclear Science & Technology)

In facilities of ITER, various construction materials are possibly exposed by tritium during periodical maintenances and an accident. It is required to establish the effective surface decontamination methods for the above construction materials of ITER. In tritium decontaminating, so-called "soaking" effect is important. This effect is based on sorption of tritiated water on the materials and subsequent desorption from them. In order to obtain and summarize data on the amount of tritium adsorption on the various materials, a series of tritiated water vapor exposure experiments have been carried out as a function of time. The amounts of tritium adsorption on the materials saturated almost within the period from several weeks to 1 month. The adsorption rate of the epoxy was found to be the largest. In the exposure time less than 2 hrs, the adsorption coefficients for the examined materials were found to be in the same order as those reported by F.Ono. It will be also discussed from viewpoint of kinetics for adsorption and desorption.

Journal Articles

Intelligible seminar on fusion reactors, 8; Fuel cycling system for tritium recovery

Fukada, Satoshi*; Hayashi, Takumi

Nippon Genshiryoku Gakkai-Shi, 47(9), p.623 - 629, 2005/09

no abstracts in English

Journal Articles

Durability of irradiated polymers in solid-polymer-electrolyte water electrolyzer

Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka; Yagi, Toshiaki; Tamada, Masao

Journal of Nuclear Science and Technology, 42(7), p.636 - 642, 2005/07

 Times Cited Count:18 Percentile:20.04(Nuclear Science & Technology)

Radioactive durability of organic polymers in solid-polymer-electrolyte water electrolyzer was investigated by $$gamma$$-ray irradiation. Serious deteriorations for tensile strength and ion exchange capacity of ion exchange membrane (Nafion) were not observed up to 850 kGy. No serious damage was also observed for the gasket materials (Aflas) up to 500 kGy. PFA and FEP, insulator materials, lost their tensile strength at 300 kGy or less. As the result, it is concluded that the electrolyzer could be used up to around 500 kGy in the case where PFA and FEP are replaced by the polyimide resin whose durability is well demonstrated. Two degrading mechanisms were supposed. One is direct degradation by energy of radial rays. The other is that by the attack of radicals. It was demonstrated that the effect of radicals on the membrane was not dominant. The quantity of dissolved fluorine in water was found to correlate with the tensile strength and ion exchange capacity. Hence, it is possible to evaluate the degradation of the membrane by monitoring the quantity of dissolved fluorine.

Journal Articles

Application of pressure swing adsorption to water detritiation process

Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka; Suzuki, Yutaka*; Kurita, Koichi*; Shimazaki, Masanori*

Journal of Nuclear Science and Technology, 42(6), p.566 - 572, 2005/06

 Times Cited Count:6 Percentile:53.71(Nuclear Science & Technology)

Pressure swing adsorption has been studied as a new water processing method for a future fusion power plant which will have a large amount of tritiated water to be processed. A series of adsorption and dehydration experiments was carried out for a typical adsorbent of NaX zeolite and it was clearly observed that break through time differs in H$$_{2}$$O and HTO, that is, it is certain that NaX zeolite can separate into the tritium concentrated water and the tritium reduced water. The quick dehydration is attained by decompression and purge gas flowing. It was observed that a part of the water released by decompression was transferred by the purge gas, and the rest water was adsorbed on the adsorbent again and was gradually released by the diffusion. It is demonstrated that enlargement of pressure difference between adsorption and dehydration is effective to obtain high dehydration ratio. Furthermore, it was also verified that enough vapor removal capacity of purge gas is quite necessary to obtain high dehydration ratio.

Journal Articles

Tritium behavoir study for detritiation of atmosphere in a room

Kobayashi, Kazuhiro; Hayashi, Takumi; Iwai, Yasunori; Asanuma, Noriko; Nishi, Masataka

Fusion Science and Technology, 41(3), p.673 - 677, 2002/05

no abstracts in English

JAEA Reports

A Study on Pore Structure of Compacted Bentonite (Kunigel-V1)

Sato, Haruo

JNC-TN8400 99-064, 22 Pages, 1999/10

JNC-TN8400-99-064.pdf:1.45MB

Four kinds of diffusion experiments; (1)through-diffusion(T-D) experiments for compaction direction dependency, (2)in-diffusion(I-D) experiments for composition dependency of silica sand in bentonite, (3)I-D experiments for initial bentonite gain size dependency, and (4)I-D experiments for the restoration property of an artificial single fracture in compacted bentonite, were carried out using tritiated water which is a non-sorbing nuclide to evaluate the effect of pore structural factors for eompacted bentonite on diffudion. For(1), effective diffusivities (De) in Na-bentonites, Kunigel-V1 and Kunipia-F were measured for 1.0 and 1.5 Mg$$cdot$$m$$^{-3}$$. For(2), apparent diffusivities (Da) in Kunigel-V1 were measured for 0.8, 1.4 and 1.8 Mg$$cdot$$m$$^{-3}$$ with silica sand of 30 and 50 wt%. For(3), Da values for 0.8, 1.4 and 1.8 Mg$$cdot$$m$$^{-3}$$ were measured for a granulated Na-bentonite, OT-9607 which grain-size distribution is in a rang between 0.1 and 5 mm. For (4), Da values in Kunigel-V1 which a single fracture was artificially reproduced and was immersed in distilled water for 7 or 28 days for the restoration of the fracture, were measured for 1.8 Mg$$cdot$$m$$^{-3}$$. Although De values in Kunigel-V1 were approximately the same for both compacted directions over the density, De values for perpendicular direction to compacted direction were higher than those for the same direction as compacted direction in Kunipia-F. For composition dependency of silica sand in bentonite, no significant effect of the mixure of silica sand in bentonite on Da was found. For initial bentonite grain size dependency, Da values obtained for OT-960 were approximately the same as those for Kunigel-V1 and no effect of initial grain size of bentonite on diffusion was found. For the restoration property of a single fracture in compacted bentonite, no restoration period dependency on Da was found. Based on this, it may be said that diffusion of nuclides in compacted bentonite, ...

JAEA Reports

None

*; Nagasaki, Shinya*

PNC-TJ1602 97-001, 57 Pages, 1997/03

PNC-TJ1602-97-001.pdf:1.2MB

None

Journal Articles

Release behavior of water from solid breeder blanket

Kawamura, Yoshinori; Okuno, Kenji; Nishikawa, Masabumi*

Proceedings of 4th International Workshop on Ceramic Breeder Blanket Interface (CBBI-4), p.235 - 248, 1995/10

no abstracts in English

JAEA Reports

Design study of fusion experimental reactor tritium plant, II; Tritiated water and radioactive solid waste processing facilities

Yoshida, Hiroshi; ; Okawa, Yoshinao; Asahara, Masaharu*; Yokogawa, Nobuhisa*; Tanemori, Nozomu*; Horikiri, Hitoshi*

JAERI-M 93-136, 117 Pages, 1993/07

JAERI-M-93-136.pdf:3.01MB

no abstracts in English

JAEA Reports

Tritium test of the tritium processing components under the Annex III US-Japan collaboration; Annex III final report

Konishi, Satoshi; Yoshida, Hiroshi; Naruse, Yuji; K.E.Binning*; R.V.Carlson*; Bartlit, J. R.*; Anderson, J. L.*

JAERI-M 93-090, 21 Pages, 1993/03

JAERI-M-93-090.pdf:0.6MB

no abstracts in English

JAEA Reports

Test of the cold traps in the JAERI fuel cleanup system in the Tritium Systems Test Assembly

Ohira, Shigeru; Konishi, Satoshi; Naruse, Yuji; Okuno, Kenji; Barnes, J. W.*; W.Harbin*; Bartlit, J. R.*; Anderson, J. L.*

JAERI-M 93-087, 29 Pages, 1993/03

JAERI-M-93-087.pdf:0.62MB

no abstracts in English

Journal Articles

Solid oxide electrolysis cell for decomposition of tritiated water

Konishi, Satoshi; ; ; Katsuta, Hiroji; Naruse, Yuji

Int.J.Hydrogen Energy, 11(8), p.507 - 512, 1986/00

 Times Cited Count:8 Percentile:24.35

no abstracts in English

JAEA Reports

In-pile Test of Tritium Release from Lithium Oxide -Tritium Release Behavior-

Kurasawa, T.; ; ; ; ;

JAERI-M 84-087, 55 Pages, 1984/05

JAERI-M-84-087.pdf:2.12MB

no abstracts in English

Journal Articles

Solld oxide electrolysis cell for decomposition of tritiated water

Konishi, Satoshi; ; ; Katsuta, Hiroji; Naruse, Yuji

Hydrogen Energy Progress,V, p.537 - 545, 1984/00

no abstracts in English

30 (Records 1-20 displayed on this page)