Noma, Yuichiro*; Kotegawa, Hisashi*; Kubo, Tetsuro*; To, Hideki*; Harima, Hisatomo*; Haga, Yoshinori; Yamamoto, Etsuji; Onuki, Yoshichika*; Ito, Kohei*; Nakamura, Ai*; et al.
Journal of the Physical Society of Japan, 90(7), p.073707_1 - 073707_5, 2021/07
Katsuta, Nagayoshi*; Ikeda, Hisashi*; Shibata, Kenji*; Kokubu, Yoko; Murakami, Takuma*; Tani, Yukinori*; Takano, Masao*; Nakamura, Toshio*; Tanaka, Atsushi*; Naito, Sayuri*; et al.
Global and Planetary Change, 164, p.11 - 26, 2018/05
Paleoenvironmental and paleoclimate changes in Siberia were reconstructed by continuous, high-resolution records of chemical compositions from a sediment core retrieved from the Buguldeika Saddle, Lake Baikal, dating back to the last 33 cal. ka BP. The Holocene climate followed by a shift at ca. 6.5 cal. ka BP toward warm and dry, suggesting that the climate system transition from the glacial to interglacial state occurred. In the last glacial period, the deposition of carbonate mud from the Primorsky Range was associated with Heinrich events (H3 and H1) and the Selenga River inflow was caused by meltwater of mountain glaciers in the Khamar-Daban Range. The anoxic bottom-water during Allerod-Younger Dryas was probably a result of weakened ventilation associated with reduced Selenga River inflow and microbial decomposition of organic matters from the Primorsky Range. The rapid decline in precipitation during the early Holocene may have been a response to the 8.2 ka cooling event.
Noma, Yuichiro*; Kotegawa, Hisashi*; Kubo, Tetsuro*; To, Hideki*; Harima, Hisatomo*; Haga, Yoshinori; Yamamoto, Etsuji; Onuki, Yoshichika*; Ito, Kohei*; Haller, E. E.*; et al.
Journal of the Physical Society of Japan, 87(3), p.033704_1 - 033704_5, 2018/03
Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*
Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06
Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.
Someya, Yoji; Tobita, Kenji; Tanigawa, Hisashi; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
This paper presents neutronics analysis mainly focused on key design issues for self-sufficient tritium production based on the conceptual design study carried out for a fusion DEMO reactor in past several years, which includes new findings regarding design methodology of breeding blanket. Self-sufficient production of tritium is one of the most critical requirements for fusion reactors. We considered a fusion DEMO reactor with a major radius of about 8 m and fusion output of 1.5 GW with breeding blanket consisting of a mixed bed of LiTiO and BeTi pebbles. The net tritium breeding ratio (TBR) was estimated to be 1.15 with a three-dimensional analysis with the MCNP-5 with nuclear library of FENDL-2.1, satisfying a self-sufficient supply of tritium (net TBR1.05). Throughout the research, we found that tritium breeding capability (i.e., local TBR) of breeding blanket is less dependent on the arrangement of cooling pipe in the blanket when the neutron wall loading is lower than about 1.5 MW/m which is met in the DEMO considered. The result suggests that tolerance for the installation of cooling pipes in each blanket module will not be a critical matter. In addition, we found that a gap of about 0.02 m between neighboring blanket modules has little impact on the gross TBR.
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
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.
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.
Nakamura, Makoto; Ibano, Kenzo*; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Ogawa, Yuichi*
Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10
Of late in Japan, a design study has been undertaken of a tokamak fusion DEMO with pressurized water coolant and solid pebble bed breeding blanket, but safety characteristics of this type of DEMO have not been well examined. In this paper, thermohydraulics analysis of in-vessel and ex-vessel loss-of-coolant accidents of a water-cooled tokamak DEMO is reported. Safety characteristics of water-cooled DEMO, particularly possible loads onto confinement barriers, are discussed based on the thermohydraulics analysis results. Measures to reduce such loads are also proposed.
Suzuki, Sadaaki; Yagyu, Junichi; Masaki, Kei; Nishiyama, Tomokazu; Nakamura, Shigetoshi; Saeki, Hisashi; Hoshi, Ryo; Sawai, Hiroaki; Hasegawa, Koichi; Arai, Takashi; et al.
NIFS-MEMO-67, p.266 - 271, 2014/02
no abstracts in English
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Tanigawa, Hisashi; Enoeda, Mikio; Tanigawa, Hiroyasu; Nakamichi, Masaru; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
This paper presents the conceptual design of a blanket with simplified structure whose interior consists of the mixture of breeder and multiplier pebble bed, cooling tubes and support for them only. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in TBR even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production. On the other hand, the thickness of blanket housing is important from the viewpoint of safety. The blanket housing may rupture when the cooling pipe in the blanket is tearing, because thickness of structure materials is thin as 22 mm. This thickness is expected to maintain to 8 MPa in the steam pressure. Finally, the blanket housing, and aspect ratio of blanket shape is proposed in consideration of TBR, and engineering problem such as maintenance and manufacture are discussed.
Matsuda, Iwao*; Nakamura, Fumitaka*; Kubo, Keisuke*; Hirahara, Toru*; Yamazaki, Shiro*; Choi, W. H.*; Yeom, H. W.*; Narita, Hisashi*; Fukaya, Yuki; Hashimoto, Mie*; et al.
Physical Review B, 82(16), p.165330_1 - 165330_6, 2010/10
no abstracts in English
Horiguchi, Hironori; Nakamura, Takemi; Kumada, Hiroaki*; Yanagie, Hironobu*; Suzuki, Minoru*; Sagawa, Hisashi
Proceedings of 14th International Congress on Neutron Capture Therapy (ICNCT-14) (CD-ROM), p.234 - 237, 2010/10
Recurrent breast cancer has been considered the application for boron neutron capture therapy using the JRR-4. The investigation of irradiation conditions for the recurrent breast cancer was performed by simulation with the JCDS. We performed the preliminary dosimetry of the model to verify the efficient irradiation conditions such as the neutron energy modes and multiple field technique. From the result, when the 30 Gy-Eq dose as minimum dose was delivered to the cancers, comparable dose distribution was delivered at the healthy tissues by both a one-port irradiation from anterior direction and a two-port irradiation from tangential direction. We revealed that the two-port irradiation was not valid to reduce the healthy tissues dose due to the isotopic scattering of neutrons in the body. We concluded that the optimal irradiation condition was the one-port irradiation with thermal neutron beam mode in terms of less healthy tissues dose and shorter irradiation time.
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.
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.
Sakanaka, Shogo*; Ago, Tomonori*; Enomoto, Atsushi*; Fukuda, Shigeki*; Furukawa, Kazuro*; Furuya, Takaaki*; Haga, Kaiichi*; Harada, Kentaro*; Hiramatsu, Shigenori*; Honda, Toru*; et al.
Proceedings of 11th European Particle Accelerator Conference (EPAC '08) (CD-ROM), p.205 - 207, 2008/06
Future synchrotron light sources based on the energy-recovery linacs (ERLs) are expected to be capable of producing super-brilliant and/or ultra-short pulses of synchrotron radiation. Our Japanese collaboration team is making efforts for realizing an ERL-based hard X-ray source. We report recent progress in our R&D efforts.
Abe, Kazuyuki; Kobayashi, Takashi*; Kajima, Hisashi*; Yoshikawa, Katsunori; Nagamine, Tsuyoshi; Nakamura, Yasuo
JAEA-Technology 2008-008, 53 Pages, 2008/03
MARICO-2 is a Testing Rig for the continuous irradiation examination of ODS ferrite steel etc.. It was necessary to re-assemble of MARICO-2 in Fuel Monitoring Facility (FMF). However, MARICO-2 is not applicable a past technology of re-assembly because it is a Rig of the total length about 11 m and its hex-tube must be welded by remote control. Then, MARICO-2 re-assembly technology development was executed, the device was designed, it produced, and the procedure of re-assembly by remote control was established.
Enoeda, Mikio; Hirose, Takanori; Tanigawa, Hisashi; Tsuru, Daigo; Suzuki, Satoshi; Ezato, Koichiro; Hatano, Toshihisa; Nomoto, Yasunobu; Nishi, Hiroshi; Homma, Takashi; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03
This paper presents progresses of the test strategy development, design and supporting R&Ds of solid breeder Test Blanket Modules (TBMs) in Japan. Japan is proposing to test its unique designs of Water Cooled Solid Breeder (WCSB) TBM and Helium Cooled Solid Breeder (HCSB) TBM from the beginning of the ITER operation. As for the design work, structure design showed steady progress and clarified detailed structure taking into account the fabrication procedure. As for supporting R&Ds, the corrosion characteristics of the structural material by high temperature and pressure water was clarified as one of critical structure integrity issues. Also, important design data of the breeder pebble bed has been clarfied. Along with the development progress, the test strategy has been investigated to obtain the most effective results of TBM test program.
Fuketa, Toyoshi; Nakamura, Takehiko; Nagase, Fumihisa; Nakamura, Jinichi; Suzuki, Motoe; Sasajima, Hideo; Sugiyama, Tomoyuki; Amaya, Masaki; Kudo, Tamotsu; Chuto, Toshinori; et al.
JAEA-Review 2006-004, 226 Pages, 2006/03
Fuel Safety Research Meeting 2005, which was organized by the Japan Atomic Energy Agency was held on March 2-3, 2005 at Toshi Center Hotel, Tokyo. The purposes of the meeting are to present and discuss the results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. The technical topics of the meeting covered the status of fuel safety research activities, fuel behavior under Reactivity Initiated Accident (RIA) and Loss of coolant accident (LOCA) conditions, high fuel behavior, and radionuclide release under severe accident conditions. This summary contains all the abstracts and sheets of viewgraph presented in the meeting.
Fukahori, Tokio; Nakamura, Hisashi*
Physical Review C, 72(6), p.064329_1 - 064329_10, 2005/12
Nuclear ground and excited state properties are described by using parameter systematics on mass and level density formulas. The formulas are based on an analytical expression of the single-particle state density by introducing the shell-paring correction in a new way. Main features are the shell, paring, and deformation effects on the droplet model near the ground state, which are washed out at higher excitation energies. The main aim of this paper is to provide in the analytical framework the improved energy dependent shell, paring, and deformation corrections generalized to the collective enhancement factors, which offer a systematic presentation over a great number of nuclear reactions. The new formulas are shown to be in close agreement with not only the empirical nuclear mass data but also the measured slow neutron resonance spacing and experimental systematics observed in the excitation energy dependent properties.