Matsuda, Shohei; Yokoyama, Keiichi; Yaita, Tsuyoshi; Kobayashi, Toru; Kaneta, Yui; Simonnet, M.; Sekiguchi, Tetsuhiro; Honda, Mitsunori; Shimojo, Kojiro; Doi, Reisuke; et al.
Science Advances (Internet), 8(20), p.eabn1991_1 - eabn1991_11, 2022/05
no abstracts in English
Sasa, Kimikazu*; Honda, Maki; Hosoya, Seiji*; Takahashi, Tsutomu*; Takano, Kenta*; Ochiai, Yuta*; Sakaguchi, Aya*; Kurita, Saori*; Satou, Yukihiko; Sueki, Keisuke*
Journal of Nuclear Science and Technology, 58(1), p.72 - 79, 2021/01
Kitazato, Kohei*; Milliken, R. E.*; Iwata, Takahiro*; Abe, Masanao*; Otake, Makiko*; Matsuura, Shuji*; Arai, Takehiko*; Nakauchi, Yusuke*; Nakamura, Tomoki*; Matsuoka, Moe*; et al.
Science, 364(6437), p.272 - 275, 2019/04
The near-Earth asteroid 162173 Ryugu, the target of Hayabusa2 sample return mission, is believed to be a primitive carbonaceous object. The Near Infrared Spectrometer (NIRS3) on Hayabusa2 acquired reflectance spectra of Ryugu's surface to provide direct measurements of the surface composition and geological context for the returned samples. A weak, narrow absorption feature centered at 2.72 micron was detected across the entire observed surface, indicating that hydroxyl (OH)-bearing minerals are ubiquitous there. The intensity of the OH feature and low albedo are similar to thermally- and/or shock-metamorphosed carbonaceous chondrite meteorites. There are few variations in the OH-band position, consistent with Ryugu being a compositionally homogeneous rubble-pile object generated from impact fragments of an undifferentiated aqueously altered parent body.
Ota, Yuki*; Sueki, Keisuke*; Sasa, Kimikazu*; Takahashi, Tsutomu*; Matsunaka, Tetsuya*; Matsumura, Masumi*; Tosaki, Yuki*; Honda, Maki*; Hosoya, Seiji*; Takano, Kenta*; et al.
JAEA-Conf 2018-002, p.99 - 102, 2019/02
no abstracts in English
Iwata, Keiko; Takamizawa, Hisashi; Ha, Yoosung; Okamoto, Yoshihiro; Shimoyama, Iwao; Honda, Mitsunori; Hanawa, Satoshi; Nishiyama, Yutaka
Photon Factory Activity Report 2017, 2 Pages, 2018/00
no abstracts in English
Ho, H. Q.; Morita, Keisuke*; Honda, Yuki; Fujimoto, Nozomu*; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Tani, Keiji*; Honda, Mitsuru; Oikawa, Toshihiro*; Shinohara, Koji; Kusama, Yoshinori; Sugie, Tatsuo
Nuclear Fusion, 55(5), p.053010_1 - 053010_15, 2015/05
The effects of a radial electric field (EF) on the losses of alpha particles and NBI fast ions in typical ITER operation scenarios for both error fields due to test blanket modules (TBMs) and toroidal field (TF) ripple were evaluated using an iterative method to execute an orbit-following Monte-Carlo code and a one-dimensional transport code. The EF effect on the loss of fast ions strongly depends on the operation scenario as well as on the error field. The electric field is very significant in the loss of fast ions in a 9MA ITER operation scenario with a higher safety factor and in the error field associated with TBMs. The EF effect in the error field of TF-ripple is very small in any operation scenario. The radial electric field changes the toroidal precession of fast ions and consequently alter their condition of resonance with the error field, which may account for the EF effect on the loss of fast ions in ITER with TBMs.
Honda, Kei; Ohira, Hiroaki; Mori, Takero
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Thermal-hydraulics analyses of the reactor upper plenum had been performed in IAEA/Monju-CRP from 2008 to 2012. However, all of the participants got a temperature distribution which didn't agree the measured data on the thermocouple plug. In this study, we re-evaluated the inlet boundary conditions and performed another analysis. The calculated temperature distribution on the thermocouple plug had good agreement with the measured data. Thermocouples and flow guide tubes are attached over the subassembly outlets. The calculated temperature at the thermocouples agreed with the temperature of the boundary conditions. And the calculated temperature at the thermocouples had good agreement with the measured data. Therefore, the temperature at the thermocouples can be regarded as the temperature of the subassembly outlets. From these results, the inlet conditions are an appropriate ones.
Mori, Takero; Sotsu, Masutake; Honda, Kei; Suzuki, Satoshi*; Ohira, Hiroaki
Journal of Energy and Power Engineering, 7(9), p.1644 - 1655, 2013/09
The prototype fast breeder reactor "MONJU" has an ex-vessel fuel storage system which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST sodium cooling system consists of three independent loops. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an station blackout (SBO) were performed. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450C. However, the structural integrity of the EVSS was maintained. The analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.
Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake
Journal of Energy and Power Engineering, 7(4), p.679 - 688, 2013/04
Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 12 Pages, 2011/09
In order to evaluate the upper plenum thermal-hydraulics of the Monju reactor vessel, we have performed detail calculations under the 40% rated power operational condition using high resolution mesh models by a commercial FVM code, FrontFlow/Red. In this study, we applied a high resolution meshes around the flow holes (FHs) on the inner barrel. We mainly made clear that the thermal-hydraulics did not change largely since the flow rates through the FHs were small enough to the total coolant flow rate but were affected largely incase without FHs on the honeycomb structure.
Honda, Kei; Ohira, Hiroaki; Sotsu, Masutake; Yoshikawa, Shinji
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10
In this study, we calculated the thermal hydraulics of the upper plenum of MONJU by the detailed analysis model using commercial FVM code, FrontFlow/Red. The present analysis model simulates all structures with high resolution meshes. The 1st order upwind and 2nd order central difference scheme were applied to the advection and diffusion terms, respectively. And RNG - model was applied to turbulence modeling. These calculation results indicated that the structures installed in the plenum except for UIS did not affect largely to the temperature and velocity, the flow characteristics in the present results had similar tendencies with porous media approached applied to the UCS region and that the difference between the temperature measured in the UCS region and that of SA outlets is relatively small.
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.
Takeuchi, Tetsuya*; Yasui, Shinichi*; Toda, Masatoshi*; Matsushita, Masaki*; Yoshiuchi, Shingo*; Oya, Masahiro*; Katayama, Keisuke*; Hirose, Yusuke*; Yoshitani, Naohisa*; Honda, Fuminori*; et al.
Journal of the Physical Society of Japan, 79(6), p.064609_1 - 064609_15, 2010/06
Yoshiuchi, Shingo*; Takeuchi, Tetsuya*; Oya, Masahiro*; Katayama, Keisuke*; Matsushita, Masaki*; Yoshitani, Naohisa*; Nishimura, Naoto*; Ota, Hisashi*; Tateiwa, Naoyuki; Yamamoto, Etsuji; et al.
Journal of the Physical Society of Japan, 79(4), p.044601_1 - 044601_11, 2010/04
Yoshiuchi, Shingo*; Toda, Masatoshi*; Matsushita, Masaki*; Yasui, Shinichi*; Hirose, Yusuke*; Oya, Masahiro*; Katayama, Keisuke*; Honda, Fuminori*; Sugiyama, Kiyohiro*; Hagiwara, Masayuki*; et al.
Journal of the Physical Society of Japan, 78(12), p.123711_1 - 123711_4, 2009/12
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).
Umeda, Koji; Oi, Takao; Osawa, Hideaki; Oyama, Takuya; Oda, Chie; Kamei, Gento; Kuji, Masayoshi*; Kurosawa, Hideki; Kobayashi, Yasushi; Sasaki, Yasuo; et al.
JAEA-Review 2007-050, 82 Pages, 2007/12
This report shows the annual report which shows the summarized results and topic outline of each project on geological disposal technology in the fiscal year of 2006.
Okajima, Satoshi*; Honda, Kei*; Sakai, Shinsuke*; Izumi, Satoshi*; Oishi, Kunio*; Kasahara, Naoto
Nihon Kikai Gakkai M&M 2007 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), p.262 - 263, 2007/10
It is well known that fatigue damage accumulates around T-junctions of piping system where two kinds of fluid with different temperatures are mixed. Kasahara et al. have proposed the thermal fatigue evaluation method based on power spectrum density (PSD). However, this method is too complex to practical use in the design stage. In this paper, a simplified damage evaluation guideline is proposed based on the Kasahara et al. method. The damage evaluation procedure from stress PSD is simplified in this guideline.
Sakai, Shinsuke*; Honda, Kei*; Okajima, Satoshi*; Izumi, Satoshi*; Kasahara, Naoto
Proceedings of 2007 ASME Pressure Vessels and Piping Division Conference/8th International Conference on Creep and Fatigue at Elevated Temperatures (PVP 2007/CREEP-8) (CD-ROM), 7 Pages, 2007/07
At an incomplete mixing area of high and low temperature fluids, fluid temperature fluctuation often occurs. It induces cyclic thermal stresses in the wall, which may result in fatigue crack initiation. Kasahara et al. proposed the thermal fatigue evaluation method based on power spectrum density (PSD). This method generalizes the evaluation procedure by preparing PSD charts of fluid and frequency transfer functions of stress for various kinds of plant components. From design point of view, however, this method is too complicated due to the inverse Fourier transform and wave decomposition procedures named Rain Flow Cycle Counting (RFC). In this paper, simplified damage evaluation method for thermal fatigue is proposed by directly evaluating fatigue damage from PSD of stress.