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Journal Articles

Development of remote pipe welding tool for divertor cassettes in JT-60SA

Hayashi, Takao; Sakurai, Shinji; Sakasai, Akira; Shibanuma, Kiyoshi; Kono, Wataru*; Onawa, Toshio*; Matsukage, Takeshi*

Fusion Engineering and Design, 101, p.180 - 185, 2015/12

 Times Cited Count:1 Percentile:12.03(Nuclear Science & Technology)

Remote pipe welding tool accessing from inside pipe has been newly developed for JT-60SA. Remote handling (RH) system is necessary for the maintenance and repair of in-vessel components such as lower divertor cassettes in JT-60SA. Cooling pipes, which connects between the divertor cassette and the vacuum vessel with bellows are required to be cut and welded in the vacuum vessel by RH system. The available space for RH system is very limited inside the vacuum vessel, especially around the divertor cassettes. Thus, the cooling pipes are required to be cut and weld from the inside in the vacuum vessel. The inner diameter, thickness and material of the cooling pipe are 54.2 mm, 2.8 mm and SUS316L, respectively. An upper pipe connected to the divertor cassette has a jut on the edge to fill the gap between pipes. Owing to the jut and two-times welding, the welding tool achieved the maximum allowable gap of 0.7 mm.

Journal Articles

Process evaluation of use of High Temperature Gas-cooled Reactors to an ironmaking system based on Active Carbon Recycling Energy System

Hayashi, Kentaro*; Kasahara, Seiji; Kurihara, Kohei*; Nakagaki, Takao*; Yan, X.; Inagaki, Yoshiyuki; Ogawa, Masuro

ISIJ International, 55(2), p.348 - 358, 2015/02

 Times Cited Count:5 Percentile:34.67(Metallurgy & Metallurgical Engineering)

Reducing coking coal consumption and CO$$_{2}$$ emissions by application of iACRES (ironmaking system based on active carbon recycling energy system) was investigated using process flow modeling to show effectiveness of HTGRs (high temperature gas-cooled reactors) adoption to iACRES. Two systems were evaluated: a SOEC (solid oxide electrolysis cell) system using CO$$_{2}$$ electrolysis and a RWGS (reverse water-gas shift reaction) system using RWGS reaction with H$$_{2}$$ produced by IS (iodine-sulfur) process. Both the effects on saving of the coking coal and reduction of CO$$_{2}$$ emissions were greater in the RWGS system. It was the reason of the result that excess H$$_{2}$$ which was not consumed in the RWGS reaction was used as reducing agent in the BF as well as CO. Heat balance in the HTGR, SOEC and RWGS modules were evaluated to clarify process components to be improved. Optimization of the SOEC temperature was desired to reduce Joule heat input for high efficiency operation of the SOEC system. Higher H$$_{2}$$ production thermal efficiency in the IS process for the RWGS system is effective for more efficient HTGR heat utilization. The SOEC system was able to utilize HTGR heat to reduce CO$$_{2}$$ emissions more efficiently by comparing CO$$_{2}$$ emissions reduction per unit heat of HTGR.

Journal Articles

Quantitative evaluation of CO$$_{2}$$ emission reduction of active carbon recycling energy system for ironmaking by modeling with Aspen Plus

Suzuki, Katsuki*; Hayashi, Kentaro*; Kurihara, Kohei*; Nakagaki, Takao*; Kasahara, Seiji

ISIJ International, 55(2), p.340 - 347, 2015/02

 Times Cited Count:7 Percentile:43.85(Metallurgy & Metallurgical Engineering)

Use of the Active Carbon Recycling Energy System in ironmaking (iACRES) has been proposed for reducing CO$$_{2}$$ emissions. To evaluate the performance of iACRES quantitatively, a process flow diagram of a blast furnace model with iACRES was developed using Aspen Plus, a chemical process simulator. CO$$_{2}$$ emission reduction and exergy analysis were performed by using mass and energy balance obtained from simulation results. The following CO$$_{2}$$ reduction methods were evaluated as iACRES: solid oxide electrolysis cells (SOEC) with CO$$_{2}$$ capture and separation (CCS), SOEC without CCS, and a reverse water-gas shift reactor powered by a high-temperature gas-cooled reactor. iACRES enabled CO$$_{2}$$ emission reduction by 3-11% by recycling CO and H$$_{2}$$, whereas effective exergy ratio decreased by 1-7%.

Journal Articles

Process evaluation of use of HTGRs to an ironmaking system based on active carbon recycling energy system (iACRES)

Hayashi, Kentaro*; Kasahara, Seiji; Kurihara, Kohei*; Nakagaki, Takao*; Yan, X.; Inagaki, Yoshiyuki; Ogawa, Masuro

Tanso Junkan Seitetsu Kenkyukai Saika Hokokusho; Tanso Junkan Seitetsu No Tenkai, p.42 - 62, 2015/02

Reducing coking coal consumption and CO$$_{2}$$ emissions by application of HTGRs (high temperature gas-cooled reactors) to iACRES (ironmaking system based on active carbon recycling energy system) was investigated using process flow modeling. Two systems were evaluated: a SOEC (solid oxide electrolysis cell) system using CO$$_{2}$$ electrolysis and a RWGS (reverse water-gas shift reaction) system using RWGS reaction with H$$_{2}$$ produced by IS (iodine-sulfur) process. Coking coal consumption was reduced from a conventional BF (blast furnace) steelmaking system by 4.3% in the SOEC system and 10.3% in the RWGS system. CO$$_{2}$$ emissions were decreased by 3.4% in the SOEC system and 8.2% in the RWGS system. Remaining H$$_{2}$$ from the RWGS reactor was used as reducing agent in the BF in the RWGS system. This was the reason of the larger reduction of coking coal consumption and CO$$_{2}$$ emissions. Electricity generation for SOEC occupied most of HTGR heat usage in the SOEC system. H$$_{2}$$ production in the IS process used most of the HTGR heat in the RWGS system. Optimization of the SOEC temperature for the SOEC system and higher H$$_{2}$$ production thermal efficiency in the IS process for the RWGS system will be useful for more efficient heat utilization. One typical-sized BF required 0.5 HTGRs and 2 HTGRs for in the SOEC system and RWGS system, respectively. CO$$_{2}$$ emissions reduction per unit heat input was larger in the SOEC system. Recycling H$$_{2}$$ to the RWGS will be useful for smaller emissions per unit heat in the RWGS system.

Journal Articles

Process modeling of iACRES by ASPEN Plus and evaluation of the whole system

Hayashi, Kentaro*; Suzuki, Katsuki*; Kurihara, Kohei*; Nakagaki, Takao*; Kasahara, Seiji

Tanso Junkan Seitetsu Kenkyukai Saika Hokokusho; Tanso Junkan Seitetsu No Tenkai, p.27 - 41, 2015/02

Applying Active Carbon Recycling Energy System to ironmaking (iACRES) process is a promising technology to reduce coal usage and CO$$_{2}$$ emissions. To evaluate performance of iACRES quantitatively, a process flow diagram of the blast furnace model with iACRES was developed using Aspen Plus. CO$$_{2}$$ emission reduction and exergy analysis was predicted by using mass and energy balance obtained from the simulation results. The followings were investigated as iACRES: solid oxide electrolysis cells (SOEC) with CO$$_{2}$$ capture and separation (CCS), SOEC without CCS, and a reverse water-gas shift reactor as the a CO$$_{2}$$ reduction reactor powered by a high-temperature gas-cooled reactor. iACRES could provide CO$$_{2}$$ emission reductions of 3-11% by recycling CO and H$$_{2}$$, whereas the effective exergy ratio decreased by 1-7%.

Journal Articles

Development of remote pipe cutting tool for divertor cassettes in JT-60SA

Hayashi, Takao; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira

Fusion Engineering and Design, 89(9-10), p.2299 - 2303, 2014/10

 Times Cited Count:6 Percentile:50.1(Nuclear Science & Technology)

Remote handling (RH) system is necessary for the maintenance and repair of in-vessel components of JT-60SA. Design study of RH system, focusing on the deployment of remote pipe cutting tool for JT-60SA divertor cassette is reported in this conference. Some cooling pipes on the outboard side in the divertor cassette should be cut and welded in the vacuum vessel. The outer diameter, thickness and material of the cooling pipe is 59.7 mm, 2.7 mm and SUS316L, respectively. Cutting tool head equips a disk cutter blade and rollers which are subjected to the reaction force. The cooling pipe is cut by rotating the cutting tool head with pushing out the disk cutter blade. Newly developed cutting tool indicates that the cooling pipe is cut by pushing out the disk cutter blade up to 30.5 mm in radius, i.e. 61 mm in diameter.

Journal Articles

Study of safety features and accident scenarios in a fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10

 Times Cited Count:10 Percentile:68.92(Nuclear Science & Technology)

After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.

Journal Articles

Key aspects of the safety study of a water-cooled fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10

Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.

Journal Articles

Carbon transport and fuel retention in JT-60U with higher temperature operation based on postmortem analysis

Yoshida, Masafumi; Tanabe, Tetsuo*; Adachi, Ayumu*; Hayashi, Takao; Nakano, Tomohide; Fukumoto, Masakatsu; Yagyu, Junichi; Miyo, Yasuhiko; Masaki, Kei; Itami, Kiyoshi

Journal of Nuclear Materials, 438, p.S1261 - S1265, 2013/07

 Times Cited Count:6 Percentile:49.64(Materials Science, Multidisciplinary)

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas in JT-60U were measured. Distributions of the fuel retention as well as the carbon re-deposition in the whole in-vessel of a large tokamak were clarified for the first time in the world. The fuel retention in the plasma shadowed areas was about two times larger than that in the carbon re-deposited layers on the plasma facing surface, although the amount of the carbon re-deposited on the plasma shadowed areas were about a half of that on the plasma facing surface, because of relatively lower temperature in the shadow areas causing higher hydrogen saturation concentration in the carbon re-deposited layers. The total fuel retention rate in JT-60U, including previously measured for all plasma facing areas, was evaluated to be 1.3$$times$$10$$^{20}$$ H+Ds$$^{-1}$$, which was lower than that in other devices, due to probably to higher temperature operation in JT-60U.

Journal Articles

Investigation of carbon dust accumulation in the JT-60U tokamak vacuum vessel

Asakura, Nobuyuki; Hayashi, Takao; Ashikawa, Naoko*; Fukumoto, Masakatsu

Journal of Nuclear Materials, 438, p.S659 - S663, 2013/07

 Times Cited Count:2 Percentile:21.27(Materials Science, Multidisciplinary)

Dust generated by the plasma-wall interaction is a potential source of the tritium retention in a fusion reactor. Dust samples were collected at 3, 5 and 3 different toroidal locations of the first wall, divertor surface and exhaust route under the divertor, respectively. On the tile surface, large number of dust was found, in particular, at the inner divertor rather upper area of the deposition layers, where recycling neutrals are increased during discharges. On the other hand, significant amount of dust (20-50 times larger) was generally accumulated at the bottom divertor, in particular, the plasma-unexposed area (remote area). It was found that the poloidal distribution is relatively symmetrical in the toroidal direction within a factor of three. Recently, analysis of dust spatial and size distributions, and evaluation of fuel retention in dust have been progressed. The total amount of the hydrogen isotope contained in the dust was estimated.

Journal Articles

Characteristics of tungsten and carbon dusts in JT-60U and evaluation of hydrogen isotope retention

Ashikawa, Naoko*; Asakura, Nobuyuki; Fukumoto, Masakatsu; Hayashi, Takao; Ueda, Yoshio*; Muroga, Takeo*

Journal of Nuclear Materials, 438, p.S664 - S667, 2013/07

 Times Cited Count:5 Percentile:43.78(Materials Science, Multidisciplinary)

In this study, W concentrations of dusts at P-8 section and hydrogen isotope retentions in dusts are analyzed. Compositions of C including W material dusts were observed in JT-60U. For an enhanced resolution of the XPS measurement to analyze quantitatively the composition of the dust flakes, the new analysis using XPS with indium foil was done and showed that the dust flakes contains about less than 1% of tungsten in carbon. Compositions of tungsten-carbon mixed dusts at different poloidal positions are reported. Produced areas of dust including W is estimated on the outer doom wing by IMPGYRO code. Relative intensities at low temperature regions were smaller than bulk divertor target, which may be caused by the high baking/operation temperature. Amounts of retained Hydrogen/Deuterium in dust particles depend on internal defects and carbon composition of dusts.

Journal Articles

Hydrogen isotopes retention in gaps at the JT-60U first wall tiles

Yoshida, Masafumi; Tanabe, Tetsuo*; Hayashi, Takao; Nakano, Tomohide; Fukumoto, Masakatsu; Yagyu, Junichi; Miyo, Yasuhiko; Masaki, Kei; Itami, Kiyoshi

Fusion Science and Technology, 63(1T), p.367 - 370, 2013/05

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

In this study, the retentions of hydrogen isotopes (H and D) in the gaps in JT-60U are clarified. Carbon tiles used in 1992-2004 were poloidally and toroidally taken out from outboard first wall in JT-60U to measure the retentions. The H and D retentions in the samples were measured by thermal desorption spectrometry (TDS). The H+D retention in the top side was higher than that of the bottom side, which might be due to thicker re-deposited carbon layers on the surface of the top side. The retentions in the surface of the side surfaces were slightly lower than that in the plasma facing surface where the retention was saturated to be 3-4e22 atoms/m$$^{2}$$. The retention rate was evaluated to be 3e17 H+D atoms/m$$^{2}$$/s from the measured retentions in two different discharge times by assuming the retention to increase linearly with the discharge time.

Journal Articles

Manufacturing and development of JT-60SA vacuum vessel and divertor

Sakasai, Akira; Masaki, Kei; Shibama, Yusuke; Sakurai, Shinji; Hayashi, Takao; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Yokoyama, Kenji; Seki, Yohji; Shibanuma, Kiyoshi; et al.

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03

The JT-60SA vacuum vessel (VV) and divertor are key components for the performance requirements. Therefore the manufacturing and development of VV and divertor are in progress, inclusive of the superconducting magnets. The vacuum vessel has a double wall structure in high rigidity to withstand electromagnetic force at disruption and to keep high toroidal one-turn resistance. In addition, the double wall structure fulfills originally two functions. (1) The remarkable reduction of the nuclear heating in the superconducting magnets is made by boric-acid water circulated in the double wall. (2) The effective baking is enabled by nitrogen gas flow of 200$$^{circ}$$C in the double wall after draining of water. Three welding types were chosen for the manufacturing of the double wall structure VV to minimize deformation by welding. Divertor cassettes with fully water cooled plasma facing components were designed to realize the JT-60SA lower single null closed divertor. The divertor cassettes in the radio-active VV have been developed to ensure compatibility with remote handling (RH) maintenance in order to allow long pulse high performance discharges with high neutron yield. The manufacturing of divertor cassettes with typical accuracy of *1 mm has been successfully completed. Brazed CFC (carbon fiber composite) monoblock targets for a divertor target have been manufactured by precise control of tolerances inside CFC blocks. The infrared thermography test of monoblock targets has been developed as new acceptance inspection.

Journal Articles

Measurements of carbon dust property in experiment and post-campaign sampling on JT-60U Tokamak

Asakura, Nobuyuki; Hayashi, Takao; Ashikawa, Naoko*; Hatae, Takaki; Nakano, Tomohide

Fusion Science and Technology, 60(4), p.1572 - 1575, 2011/11

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

Dust research has been performed in JT-60U in order to predict the plasma performance and the tritium retention in a fusion reactor. Laser scattering measurement showed, in specific discharge after disruptions, both the size and number were peaked in the far-SOL and they decreased near the separatrix. This result shows that sublimation of dust is dominant in the SOL. Dust collection after the experiment campaign showed that large weight of the dust was cumulated on the exhaust route of gas flow under the divertor. The total amount of the hydrogen isotope contained in the dust were estimated for the cases with deposited in the volume and near the surface.

Journal Articles

Measurement of dust quantity and distribution collected from JT-60U

Hayashi, Takao; Asakura, Nobuyuki; Ashikawa, Naoko*; Nakano, Tomohide

Fusion Science and Technology, 60(4), p.1548 - 1551, 2011/11

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

A real mass densities of carbon dust collected in the baffle and divertor regions of JT-60U were investigated. On the plasma-exposed surface, large areal density of 610 mg/m$$^{2}$$ is found at the upper tile of the inner divertor, which is much larger than other areas due to the soft deposition. On the other hand, as for the plasma-shadowed area, largest areal density of 5,100 mg/m$$^{2}$$ was found underneath the dome structure. The total dust weights at the plasma-exposed surface and the shadowed areas were estimated to be 1.3 g and 22.2 g, respectively, assuming the toroidal symmetry. Count-based size distributions were also investigated. The average dust size of the main population is less than 20 $$mu$$m for both the plasma-exposed surface and the shadowed area.

Journal Articles

Design study of remote handling system for lower divertor cassettes in JT-60SA

Hayashi, Takao; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira

Fusion Science and Technology, 60(2), p.549 - 553, 2011/08

 Times Cited Count:4 Percentile:36.36(Nuclear Science & Technology)

Design study of RH system, especially the expansion of the RH rail and replacement of the lower divertor cassettes, was described in this paper. The dimensions and weight of the divertor cassette, which is 10 degrees wide in toroidal direction, are 1.62$$^{L}$$ $$times$$ 0.57$$^{W}$$ $$times$$ 1.25$$^{H}$$ m and 800 kg, respectively. The RH system can use four horizontal ports whose inside dimensions are 0.66$$^{W}$$ $$times$$ 1.83$$^{H}$$ m. The space for RH system is very limited. The RH rail is installed before transporting divertor cassettes. The RH rail can cover 180 degrees in toroidal direction. A divertor cassette can be replaced by heavy weight manipulator (HWM) consists of an end effector, a telescopic arm and a vehicle. The HWM brings the divertor cassette to the front of another horizontal port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device, which is installed from outside the vacuum vessel, receives and brings out the divertor cassette.

JAEA Reports

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m$$^{2}$$. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Journal Articles

Design of lower divertor for JT-60SA

Sakurai, Shinji; Higashijima, Satoru; Hayashi, Takao; Shibama, Yusuke; Masuo, Hiroshige*; Ozaki, Hidetsugu; Sakasai, Akira; Shibanuma, Kiyoshi

Fusion Engineering and Design, 85(10-12), p.2187 - 2191, 2010/08

 Times Cited Count:9 Percentile:53(Nuclear Science & Technology)

JT-60SA tokamak project has just started construction phase under both the Japanese domestic program and the Japan-EU international program "ITER Broader Approach". All of plasma facing components (PFC) shall be actively cooled due to high power long pulse plasma heating. Lower single null closed divertor with vertical target (VT) will be installed at the start of experiment phase. Each divertor module covers a 10-degree sector in toroidal direction. PFCs such as VTs, baffles and dome shall be assembled on a divertor cassette, which provides integrated coolant pipe connection to coolant headers in the VV. Static structural analysis for dead weight, coolant pressure and EM loads shows that displacement and stress of the divertor module are generally small but a part of support structure of PFC requires improvement.

Journal Articles

Analysis of residual gas by high-resolution mass spectrometry during helium glow discharge cleaning in JT-60U

Hayashi, Takao; Kaminaga, Atsushi; Arai, Takashi; Sato, Masayasu

Fusion Engineering and Design, 84(2-6), p.908 - 910, 2009/06

 Times Cited Count:3 Percentile:26.15(Nuclear Science & Technology)

The residual gas analysis has been conducted by high-resolution mass spectrometry which can discriminate between D$$_{2}$$ and He gas species during helium glow discharge cleaning (He-GDC) in JT-60U in order to investigate the effect of He-GDC. The residual gas analyzer was able to distinguish between D$$_{2}$$ and He peaks during He-GDC. Since the He-GDC started, the partial pressure of D$$_{2}$$ gas increases with time and reached its highest pressure (3.8 $$times$$ 10$$^{-4}$$ Pa), which is about ten times larger than that before the He-GDC (3.5 $$times$$ 10$$^{-5}$$ Pa). The total amount of D$$_{2}$$, which was released during the He-GDC (7 hours), was evaluated as 4 Pa m$$^{3}$$. The pressure of D$$_{2}$$ (5.7 $$times$$ 10$$^{-6}$$ Pa) about 7 hours after the He-GDC (7 hours) is significantly lower than before the He-GDC, which indicates the He-GDC is effective to remove the deuterium from plasma facing components.

Journal Articles

Hydrogen isotope retention in the outboard first wall tiles of JT-60U

Yoshida, Masafumi; Tanabe, Tetsuo*; Nobuta, Yuji*; Hayashi, Takao; Masaki, Kei; Sato, Masayasu

Journal of Nuclear Materials, 390-391, p.635 - 638, 2009/06

 Times Cited Count:9 Percentile:56.98(Materials Science, Multidisciplinary)

We have investigated hydrogen isotopes retention in the outboard first wall tiles of JT-60U by means of TDS, SIMS and SEM. The outboard first wall tiles of JT-60U are mostly eroded. The total retention (H+D) normalized by a unit area and integrated NBI time in the eroded first wall tiles and the eroded divertor tiles were nearly the same, in spite of the lower temperature of the first wall. Differently from divertor tiles, in which H retention was dominated owing to HH discharges preformed after DD discharges, deuterium is dominated in hydrogen isotopes retention and penetrates deeper from the surface. This is attributed to injection of high energy D and difficulty of isotopic replacement owing to their lower temperature. The integrated amount over the whole surface could be appreciably large. This type of hydrogen retention could be also possible for the metallic wall.

84 (Records 1-20 displayed on this page)