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Journal Articles

Tritium release behavior from steels irradiated by high energy protons

Nakamura, Hirofumi; Kobayashi, Kazuhiro; Yamanishi, Toshihiko; Yokoyama, Sumi; Saito, Shigeru; Kikuchi, Kenji

Fusion Science and Technology, 52(4), p.1012 - 1016, 2007/11

 Times Cited Count:2 Percentile:19.85(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Recent progress in solid breeder blanket development at JAEA

Nishitani, Takeo; Enoeda, Mikio; Akiba, Masato; Yamanishi, Toshihiko; Hayashi, Kimio; Tanigawa, Hiroyasu

Fusion Science and Technology, 52(4), p.971 - 978, 2007/11

 Times Cited Count:9 Percentile:57.83(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Progress in neutronics studies for the water cooled pebble bed blanket

Nishitani, Takeo; Sato, Satoshi; Ochiai, Kentaro; Kutsukake, Chuzo; Tanaka, Shigeru; Abe, Yuichi; Konno, Chikara

Fusion Science and Technology, 52(4), p.791 - 795, 2007/11

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

no abstracts in English

Journal Articles

Spectral effects of activation for liquid blanket relevant materials induced by D-T neutron irradiation

Li, Z.*; Tanaka, Teruya*; Muroga, Takeo*; Sato, Satoshi; Nishitani, Takeo

Fusion Science and Technology, 52(4), p.817 - 820, 2007/11

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

no abstracts in English

Journal Articles

Design progress of the ITER in-wall shielding

Morimoto, Masaaki; Ioki, Kimihiro; Terasawa, Atsumi; Utin, Y.*

Fusion Science and Technology, 52(4), p.834 - 838, 2007/11

 Times Cited Count:1 Percentile:11.92(Nuclear Science & Technology)

The ITER in-wall shielding is mounted in between the double walls of the Vacuum Vessel. Boron-doped stainless steel and SS430 ferritic steel are used. The design improvement of the in-wall shielding has focused on reducing electromagnetic forces acting on shielding blocks. It has been found that the calculated electromagnetic forces have been significantly reduced. Magnetization forces have also been calculated for ferromagnetic inserts. Based on these load conditions, structural analyses have been performed and structural integrity has been validated. Shapes of boron-doped shielding plates which have low ductility are carefully designed to prevent excessive stress concentrations and not to take high mechanical loads. This makes shielding plate design simpler and more robust. Suitable dimensions and gaps between shielding blocks and between shielding block and the VV have been designed to fit to tolerances of the VV.

Journal Articles

Effects of tube drawing on structural material for ITER test blanket module

Hirose, Takanori; Tanigawa, Hiroyasu; Enoeda, Mikio; Akiba, Masato

Fusion Science and Technology, 52(4), p.839 - 843, 2007/11

 Times Cited Count:9 Percentile:57.83(Nuclear Science & Technology)

This paper presents the effects of tube milling process on microstructural properties of F82H, which is one of the most important issues for fabrication of the module. The 3500 mm long square tubes have been developed by cold rolling method. This tube is long enough to fabricate the first wall without any joint in the cooling path. Its surface roughness (Rz) and outer curvature are less than 1 microns and 1.4 mm, respectively. It is fine enough to reduce the assembly gap for Hot Isostatic Pressing (HIP) joint. Although the process introduced stretched microstructure containing dense precipitates, this anisotropic microstructure was successfully recovered by heat treatments corresponding to HIP process. This shows the milling process verified in this work could be applicable to blanket fabrication process.

Journal Articles

Activation analysis for sequential reactions of a fusion Demo-reactor

Yamauchi, Michinori; Nishitani, Takeo; Nishio, Satoshi; Hori, Junichi*; Kawasaki, Hiromitsu*

Fusion Science and Technology, 52(4), p.781 - 785, 2007/11

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

Low activation material is one of the important factors for constructing high power fusion reactors in future. Unexpected activation, however, may be produced through sequential reactions due to charged particles created by primary neutron reactions. In the present work, the effect of the sequential activation reaction was studied for candidate low activation materials of a fusion demo-reactor. The calculations were conducted by the ACT4 code developed in JAEA for the activation analysis of fusion reactor designs and revised for dealing with the sequential activation reactions. The results say that the real dose rate around vanadium alloy which may be used as structural material becomes larger after the cooling for 3 years by considering the reaction. Although metal hydrate is regarded as an excellent low activation shield material, the reactions due to recoil protons are influential and the dose rate around vanadium hydrate is several orders of magnitude larger than the value calculated without the sequential process after 2 weeks cooling. In case of liquid breeders, the effect of sequential reactions is popularly observed and it affects the shield design of circulation loop.

Journal Articles

Overview of recent Japanese activities and plans in fusion technology

Yamamoto, Ichiro*; Nishitani, Takeo; Sagara, Akio*

Fusion Science and Technology, 52(3), p.347 - 356, 2007/10

 Times Cited Count:2 Percentile:19.85(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Tritium behavior on the water-metal boundary for the permeation into cooling water through metal piping

Hayashi, Takumi; Nakamura, Hirofumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Yamanishi, Toshihiko; Okuno, Kenji*

Fusion Science and Technology, 52(3), p.687 - 691, 2007/10

 Times Cited Count:10 Percentile:60.87(Nuclear Science & Technology)

How to confine tritium within high temperature breeding blanket is the key issue for safety and fuel economy of the fusion reactor. Specially, tritium permeation into cooling water is very important, however, there is little report of the systematic experiment comparing with that into gaseous coolant. Therefore, a series of tritium transportation experiments into water was performed through pure iron piping samples, which contained more than 1 kPa of pure tritium gas and fixed inside the water jacket under controlled temperature and pressure. Chemical species of tritium in water were measured during the experiment until reaching enough stable permeation, and tritium distribution/situation on the metal surface layer was also measured using autoradiography etc. after the experiment. In this paper, the results of tritium transportation experiments were summarized and tritium behavior on the boundary surface between metal piping and cooling water was discussed.

Journal Articles

Recent activities on tritium technologies for ITER and fusion reactors at JAEA

Hayashi, Takumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Iwai, Yasunori; Kawamura, Yoshinori; Nakamura, Hirofumi; Shu, Wataru; Arita, Tadaaki; Hoshi, Shuichi; Suzuki, Takumi; et al.

Fusion Science and Technology, 52(3), p.651 - 658, 2007/10

 Times Cited Count:2 Percentile:19.85(Nuclear Science & Technology)

The design studies of Air Detirtiation 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 R&D of the tritium recovery system of ITER test blanket, using mainly molecular sieve and/or electro-chemical pumping system. A series of fundamental studies on tritium safety technologies, such as tritium behavior in a confinement and its barrier materials, monitoring, accountancy, detritiation and decontamination etc., has been carried out as a major activity in JAEA for ITER and fusion demo reactors. In this paper, the above recent activities on tritium technologies at Tritium Process Lab. in JAEA are summarized for ITER and future fusion reactor.

Journal Articles

Self-decomposition behavior of high concentration tritiated water

Ito, Takeshi*; Hayashi, Takumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Yamanishi, Toshihiko

Fusion Science and Technology, 52(3), p.701 - 705, 2007/10

 Times Cited Count:1 Percentile:11.92(Nuclear Science & Technology)

In a fusion reactor, how to handle high concentration tritiated water (HTO) is one of the key issues for the safety control. High concentration HTO decomposes by itself, and generates hydrogen and oxygen in the gas phase and hydrogen peroxide in the liquid phase, as the final products. There are many report of the G values for water decomposition by $$gamma$$-ray irradiation experiments, however, those for self-decomposition of HTO are limited because of the difficulty of the safety handling of HTO. In the Tritium Process Lab. of Japan Atomic Energy Agency, the characteristics of a wide rage of HTO up to about 2 EBq/m$$^{3}$$ has been investigated for more than 10 years. The effective G values of hydrogen and hydrogen peroxide under self-decomposition of HTO were evaluated from their concentration increase in the leak tight vessel stored HTO, as a function of time. In this paper, the above effective G values are summarized, and the dependences of HTO concentration and temperature are discussed.

Journal Articles

Isotope effect of hydrogen rapidly supplied from the metal storage bed

Hayashi, Takumi; Suzuki, Takumi; Shu, Wataru; Yamanishi, Toshihiko

Fusion Science and Technology, 52(3), p.706 - 710, 2007/10

 Times Cited Count:8 Percentile:53.84(Nuclear Science & Technology)

In the tritium Storage and Delivery System (SDS) of ITER, how to control the isotope balance of DT fuel is one of the key issues for the stable and optimum operation. Basically, the equilibrium pressure of hydrogen-metal system has large isotope effect such as PH$$_{2}$$ $$<$$ PD$$_{2}$$ $$<$$ PT$$_{2}$$, however, there is only a limited data of the isotope composition of hydrogen mixture, which is supplied rapidly from the storage bed by a vacuum pump under ITER/SDS conditions. Therefore, in order to investigate the isotope composition of supplied hydrogen gases, a series of rapid supply experiments was performed using a 1/10 ITER scale ZrCo bed with a scroll pump as functions of bed temperature (573 K $$sim$$ 623 K) and isotope composition of hydrogen mixture stored initially (H:D = 1:9 $$sim$$ 9:1). The isotope composition was measured by in-line mass spectrometer during continuous hydrogen supply. In this paper, the above results are summarized and the isotope effect is discussed. The effective way to control the isotope balance of DT fuel is also discussed with more moderate SDS design conditions.

Journal Articles

Superconducting tokamak JT-60SA project for ITER and DEMO researches

Hosogane, Nobuyuki; JT-60SA Design Team; JA-EU Satellite Tokamak Working Group

Fusion Science and Technology, 52(3), p.375 - 382, 2007/10

 Times Cited Count:6 Percentile:44.88(Nuclear Science & Technology)

JT-60SA is a superconducting tokamak with wide flexibility in plasma shape and single/double null divertors, capable of confining break-even class high temperature plasma for 100 s with intensive heating power of 41 MW. The design of JT-60SA is based on the National Centralized Tokamak NCT, and has been proposed as an ITER satellite tokamak in the 10 years Broader Approach (BA) program between Japan and Europe. The JT-60SA project is a combination of the BA project and the NCT project. The former mission is to support ITER by developing an understanding of physics issues, optimizing operation scenarios etc.. The latter mission is mainly to explore steady state, high beta DEMO relevant scenarios. The construction period is 7 years and 3 years will be devoted to experimental studies with a possibility of extension. The detail design of JT-60SA is progressing under their collaboration. Overview of the project and machine design is presented.

Journal Articles

Study for the behavior of tritiated water vapor on organic materials

Kobayashi, Kazuhiro; Hayashi, Takumi; Nakamura, Hirofumi; Yamanishi, Toshihiko; Oya, Yasuhisa*; Okuno, Kenji*

Fusion Science and Technology, 52(3), p.696 - 700, 2007/10

 Times Cited Count:1 Percentile:11.92(Nuclear Science & Technology)

In a fusion reactor of high safety and acceptability, safe confinement of tritium is one of key issues for the fusion reactor. Tritium should be well-controlled and not excessively released to the environment and to prevent workers from excess exposure. Especially, the hot cell and tritium facilities of ITER will use various construction materials. For tritium decontamination processes, so-called soaking effect is important. This effect is based on sorption of tritiated water vapor on the materials and subsequent desorption from them. Therefore, in order to develop for the optimal decontamination technique, the decontamination experiment was carried out as a function of water vapor concentration in the purging gas for epoxy paint, acrylic resin and butyl rubber. As the result, about 70% of the adsorbed tritium on the epoxy paint was removed by adding water vapor in purging gas for 12 hrs. The effect of adding water vapor was found on the decontamination for epoxy paint.

Journal Articles

Oxidation performance test of detritiation system under existence of SF$$_{6}$$

Kobayashi, Kazuhiro; Miura, Hidenori*; Hayashi, Takumi; Hoshi, Shuichi; Yamanishi, Toshihiko

Fusion Science and Technology, 52(3), p.711 - 715, 2007/10

 Times Cited Count:3 Percentile:26.87(Nuclear Science & Technology)

The tritium released in the building is removed by Atmosphere Detritiation System (ADS), where the tritium is oxidized by catalysts and is removed as water. Special gas of SF$$_{6}$$ is used as an electric insulation gas in ITER, and is expected to be released in an accident such as fire. Although SF$$_{6}$$ has the potential as a catalyst poison, the performance of ADS with the existence of SF$$_{6}$$ has not been confirmed yet. Therefore, to study the effect of SF$$_{6}$$, the performance tests of ADS was carried out with air containing $$sim$$1% of hydrogen, $$sim$$1% of methane and $$sim$$1% of SF$$_{6}$$. The SF$$_{6}$$ gas was notably decomposed in the case of the catalyst temperature higher than 673 K. In addition, a part of the water produced by the 473 K catalyst was reduced to hydrogen due to the reaction with the decomposed gas in SF$$_{6}$$. Consequently, the detritiation factor of ADS was decreased to less than 50 from $$>$$ 1000 of its initial value.

Journal Articles

Hydrogen isotope retentions and erosion/deposition profiles in the first wall of JT-60U

Oya, Yasuhisa*; Hirohata, Yuko*; Nakahata, Toshihiko*; Suda, Taichi*; Yoshida, Masashi*; Arai, Takashi; Masaki, Kei; Okuno, Kenji*; Tanabe, Tetsuo*

Fusion Science and Technology, 52(3), p.554 - 558, 2007/10

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

To investigate retention characteristics of hydrogen isotopes in the first wall tiles of JT-60U, surface morphology, erosion/deposition profiles and hydrogen isotope retentions were examined by SEM, XPS, TDS and SIMS. It was found that poloidal deuterium retention profile was rather uniform, while the thermal desorption behavior of deuterium was quite different depending on the locations of the tiles. Deuterium retained in the upper first wall, where was covered by thick boron layers with high concentration of B, was desorbed at lower temperature than that in the lower area covered by carbon layers with much less B content. D/H ratio in the first wall tiles was appreciably higher than that observed in the divertor tiles, suggesting the injection of high energy deuteron originating from NBI into the first wall. In addition, the lower temperature of the first wall compared to that of the divertor tiles would prohibit desorption of the implanted deuterium and/or its replacement by subsequent D or H impingement.

Journal Articles

Fundamental study on purity control of the liquid metal blanket using solid electrolyte cell

Yamamoto, Yoshihiko*; Yamanishi, Toshihiko; Kawamura, Yoshinori; Isobe, Kanetsugu; Yamamoto, Yasushi*; Konishi, Satoshi*

Fusion Science and Technology, 52(3), p.692 - 695, 2007/10

 Times Cited Count:1 Percentile:11.92(Nuclear Science & Technology)

A solid electrolyte (ionic conductor) cell has been developed for measurement and control of hydrogen and oxygen density in liquid metal blanket with LiPb. The ceramic tubes of the SrCe$$_{0.95}$$Yb$$_{0.05}$$O$$_{3-x}$$ (proton conductor) and Yttria Stabilized Zirconia (oxygen ion conductor) that can be used at operating temperature of LiPb blanket have been utilized for electrolytes of the devices, which enable the continuous measurement of respectively hydrogen and oxygen in liquid LiPb. Fundamental electrochemical data such as EMF for a partial pressure and ion conductivity were measured, and the results were used to evaluate the feasibility of these devices.

Oral presentation

Light elements behavior in fusion devices studied with JT-60 and oxygen discharge

Konishi, Satoshi*; Nakamura, Hirofumi; Isobe, Kanetsugu; Yamamoto, Yasushi*; Kaminaga, Atsushi

no journal, , 

In order to understand the behavior of hydrogen isotopes, carbon, and oxygen in the fusion devices, and to develop effective method to control codeposite and tritium inventory, exhaust gas from JT-60 was analyzed. Unsaturated hydrocarbons such as acetylene was found in tokamak exhaust, suggesting relation with codeposite formation. Oxygen-helium glow discharge in JT-60 indicated enhanced carbon removal corresponding to sputtering yield of 2.5 for carbon per oxygen ion. Simulating discharge experiment in a small chamber with carbon electrodes also reproduced codeposition by acetylene discharge and its removal by O$$_{2}$$-He glow.

Oral presentation

Conceptual design of advanced blanket using liquid Li-Pb

Ueno, Yukihisa*; Niigawa, Satoshi*; Enoeda, Mikio; Yamamoto, Yasushi*; Konishi, Satoshi*

no journal, , 

no abstracts in English

Oral presentation

Evaluation of advanced materials for liquid Li-Pb blanket environment

Niigawa, Satoshi*; Ueno, Yukihisa*; Enoeda, Mikio; Hinoki, Tatsuya*; Park, J.*; Yamamoto, Yasushi*; Konishi, Satoshi*

no journal, , 

no abstracts in English

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