Hayashi, Kentaro*; Kasahara, Seiji; Kurihara, Kohei*; Nakagaki, Takao*; Yan, X.; Inagaki, Yoshiyuki; Ogawa, Masuro
ISIJ International, 55(2), p.348 - 358, 2015/02
Reducing coking coal consumption and CO 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 electrolysis and a RWGS (reverse water-gas shift reaction) system using RWGS reaction with H produced by IS (iodine-sulfur) process. Both the effects on saving of the coking coal and reduction of CO emissions were greater in the RWGS system. It was the reason of the result that excess H 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 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 emissions more efficiently by comparing CO emissions reduction per unit heat of HTGR.
Suzuki, Katsuki*; Hayashi, Kentaro*; Kurihara, Kohei*; Nakagaki, Takao*; Kasahara, Seiji
ISIJ International, 55(2), p.340 - 347, 2015/02
Use of the Active Carbon Recycling Energy System in ironmaking (iACRES) has been proposed for reducing CO 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 emission reduction and exergy analysis were performed by using mass and energy balance obtained from simulation results. The following CO reduction methods were evaluated as iACRES: solid oxide electrolysis cells (SOEC) with CO 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 emission reduction by 3-11% by recycling CO and H, whereas effective exergy ratio decreased by 1-7%.
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 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 electrolysis and a RWGS (reverse water-gas shift reaction) system using RWGS reaction with H 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 emissions were decreased by 3.4% in the SOEC system and 8.2% in the RWGS system. Remaining H 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 emissions. Electricity generation for SOEC occupied most of HTGR heat usage in the SOEC system. H 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 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 emissions reduction per unit heat input was larger in the SOEC system. Recycling H to the RWGS will be useful for smaller emissions per unit heat in the RWGS 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 emissions. To evaluate performance of iACRES quantitatively, a process flow diagram of the blast furnace model with iACRES was developed using Aspen Plus. CO 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 capture and separation (CCS), SOEC without CCS, and a reverse water-gas shift reactor as the a CO reduction reactor powered by a high-temperature gas-cooled reactor. iACRES could provide CO emission reductions of 3-11% by recycling CO and H, whereas the effective exergy ratio decreased by 1-7%.
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.
Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10
The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.
Hayashi, Kentaro*; Katayanagi, Nobuko*; Fumoto, Tamon*; Hasegawa, Toshihiro*; Ono, Keisuke*; Katata, Genki
Nogyo Kankyo Gijutsu Kenkyojo Heisei 25 Nendo Kenkyu Seika Joho, 30 (Internet), 2 Pages, 2014/03
no abstracts in English
Katata, Genki; Hayashi, Kentaro*; Ono, Keisuke*; Nagai, Haruyasu; Miyata, Akira*; Mano, Masayoshi*
Agricultural and Forest Meteorology, 180, p.1 - 21, 2013/10
A multi-layer atmosphere-SOiL-VEGetation model (SOLVEG) was modified to calculate the NH exchange fluxes over a paddy field. The heat transfer at the paddy water layer and the dry deposition of water-soluble gases such as NH and SO onto the wet canopy, as well as the emission potentials of NH from the rice foliage and the surface of floodwater or soil were incorporated into the model. The modified model reproduced the observed surface and NH fluxes, paddy water temperature, and soil temperature and moisture during both the fallow and cropping seasons. The "recaptured fraction" was defined as the ratio of the amount of volatilized NH recaptured by the foliage to the total amount. Numerical experiments using the modified model with varying emission potentials of NH showed that the recaptured fraction increased with an increase in the leaf area index (LAI) and saturated when LAI 1 because of the limitation of stomatal uptake.
Kawamura, Yoshinori; Ochiai, Kentaro; Hoshino, Tsuyoshi; Kondo, Keitaro*; Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamichi, Masaru; Konno, Chikara; Yamanishi, Toshihiko; Hayashi, Takumi; et al.
Fusion Engineering and Design, 87(7-8), p.1253 - 1257, 2012/08
Tritium generation and recovery study on lithium ceramic packed bed was started by use of FNS in JAEA. Lithium titanate was selected as tritium breeding material. In this work, the effect of sweep gas species on tritium release behavior was investigated. In case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in case of sweep by dry helium, tritium in gaseous form was released first, and release of tritium in water form was delayed and was gradually increased.
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
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. 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.
Morita, Takami*; Niwa, Kentaro*; Fujimoto, Ken*; Kasai, Hiromi*; Yamada, Haruya*; Nishiuchi, Ko*; Sakamoto, Tatsuya*; Godo, Waichiro*; Taino, Seiya*; Hayashi, Yoshihiro*; et al.
Science of the Total Environment, 408(16), p.3443 - 3447, 2010/06
Iodine-131 (I) was detected in brown algae collected off the Japanese coast. The maximum measured specific activity of I in brown algae was 0.370.010 Bq/kg-wet. Cesium-137 (Cs) was also detected in all brown algal samples used in this study. There was no correlation between specific activities of I and Cs in these seaweeds. Low specific activity and minimal variability of Cs in brown algae indicated that past nuclear weapon tests were the source of Cs. Although nuclear power facilities are known to be pollution sources of I, there was no relationship between the sites where I was detected and the locations of nuclear power facilities. Most of the sites where I was detected were near big cities with large populations. On the basis of the results, we suggest that the likely pollution source of I, detected in brown seaweeds, is not nuclear power facilities, but nuclear medicine procedures.
Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.
Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07
Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).
Tanigawa, Hisashi; Hoshino, Tsuyoshi; Kawamura, Yoshinori; Nakamichi, Masaru; Ochiai, Kentaro; Akiba, Masato; Ando, Masami; Enoeda, Mikio; Ezato, Koichiro; Hayashi, Kimio; et al.
Nuclear Fusion, 49(5), p.055021_1 - 055021_6, 2009/05
This paper presents recent achievements of the research activities for the TBM being developed in JAEA, focusing on the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of LiTiO has been improved by LiO additives. In order to analyze the pebble bed behaviour, thermo-mechanical properties of the LiTiO pebble bed has been experimentally obtained. In order to verify nuclear properties of the pebble bed, the activation foil method has been proposed and a preliminary experiment has been conducted. For the tritium behaviour, the chemical densified coating method has been well developed and tritium recovery system has been modified taking account of the design change of the TBM.
Kubota, Naoyoshi; Ochiai, Kentaro; Kutsukake, Chuzo; Hayashi, Takao; Shu, Wataru; Kondo, Keitaro; Verzilov, Y.*; Sato, Satoshi; Yamauchi, Michinori; Nishi, Masataka; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 7 Pages, 2007/03
Fuel and impurity particles show complicated behavior on the surface of plasma facing components (PFC) in fusion devices. The study is important for the design of the fuel recycling, safety management of the tritium inventory, etc. Quantitative measurements of hydrogen and lithium isotopes together with other impurities on the PFC surface exposed to D-T plasmas in TFTR were performed using the deuteron-induced nuclear reaction analysis, imaging plate method, full combustion method and activation analysis. The tritium depth profile was different from deuterium one. The surface tritium largely contributed to the whole tritium in the sample. On the other hand, the retained amount of lithium-6 was lager than that of lithium-7. This relates to the injection of enriched lithium-6 pellets in some campaigns. No other impurities were detected. So the large amount of tritium remained near the surface and did not diffuse more deeply, which gives a bright prospect for tritium safety.
Hayashi, Takao; Ochiai, Kentaro; Masaki, Kei; Goto, Yoshitaka*; Kutsukake, Chuzo; Arai, Takashi; Nishitani, Takeo; Miya, Naoyuki
Journal of Nuclear Materials, 349(1-2), p.6 - 16, 2006/02
Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JT-60U were investigated by NRA. The highest deuterium concentration of D/C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium-carbon co-deposition. In the outer and inner divertor target tiles, the D/C data were lower than 0.006. Additionally, the maximum (H+D)/C in the dome top tile was estimated to be 0.023 from the results of NRA and SIMS. OFMC simulation showed energetic deuterons caused by NBI were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, SEM observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium-carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk.
Suzuki, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Hirose, Takanori; Hayashi, Kimio; Tanigawa, Hisashi; Ochiai, Kentaro; Nishitani, Takeo; Tobita, Kenji; Akiba, Masato
Nuclear Fusion, 46(2), p.285 - 290, 2006/02
This paper presents significant progress in R&D of key technologies on the water-cooled solid breeder blanket for the ITER-TBM in JAERI. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 C followed by normalizing at 930 C after the HIP process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a uniaxial hot compression without any artificial compliant layer. Also, it has been confirmed that a fatigue lifetime correlation, which was developed for ITER divertor, can be applicable for F82H first wall mock-up. As for R&D on a breeder material, LiTiO, the effect of compression loads on thermal conductivity of pebble beds has been clarified. JAERI have extensively developed key technologies for ITER-TBM, and now steps up into an engineering R&D stage, where integrated performance of TBM structures will be demonstrated by scalable mock-ups.
Enoeda, Mikio; Hatano, Toshihisa; Tsuchiya, Kunihiko; Ochiai, Kentaro; Kawamura, Yoshinori; Hayashi, Kimio; Nishitani, Takeo; Nishi, Masataka; Akiba, Masato
Fusion Science and Technology, 47(4), p.1060 - 1067, 2005/05
Japan Atomic Energy Research Institute (JAERI) has been assigned as a leading institute for developing the solid breeder blanket in the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. In accordance with the long term research program, element technology development of solid blanket has been performed at JAERI and showed significant progress. Based on the achievement of the element technology development, the development phase is now stepping further to the engineering development phase. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI.
Kubota, Naoyoshi; Ochiai, Kentaro; Kutsukake, Chuzo; Hayashi, Takao; Shu, Wataru; Nishi, Masataka; Nishitani, Takeo
Purazuma, Kaku Yugo Gakkai-Shi, 81(4), p.296 - 301, 2005/04
no abstracts in English
Masaki, Kei; Sugiyama, Kazuyoshi*; Hayashi, Takao; Ochiai, Kentaro; Goto, Yoshitaka*; Shibahara, Takahiro*; Hirohata, Yuko*; Oya, Yasuhisa*; Miya, Naoyuki; Tanabe, Tetsuo*
Journal of Nuclear Materials, 337-339, p.553 - 559, 2005/03
no abstracts in English
Ochiai, Kentaro; Hayashi, Takao; Kutsukake, Chuzo; Goto, Yoshitaka*; Masaki, Kei; Arai, Takashi; Miya, Naoyuki; Nishitani, Takeo
Journal of Nuclear Materials, 329-333(Part1), p.836 - 839, 2004/08
no abstracts in English
Morioka, Atsuhiko; Sato, Satoshi; Kinno, Masaharu*; Sakasai, Akira; Hori, Junichi*; Ochiai, Kentaro; Yamauchi, Michinori*; Nishitani, Takeo; Kaminaga, Atsushi; Masaki, Kei; et al.
Journal of Nuclear Materials, 329-333(2), p.1619 - 1623, 2004/08
The neutron penetration and the activation characteristics of the boron-doped low activation concrete were investigated for irradiation of 2.45 and 14 MeV neutrons. The shielding property of the 2 wt% boron-doped low activation concrete is superior to that of the 1 wt% boron for the thermal neutron, on the contrary to the no clear difference for the fast neutron. The total activity detected in the boron-doped low activation concrete was about one hundredth of that in the geostandard sample at more than 30 days cooling time. The total activity of the boron-doped concrete by major nuclei does not depend on the boron density for the 14 MeV neutron irradiation.