Sakuma, Kazuyuki; Tsuji, Hideki*; Hayashi, Seiji*; Funaki, Hironori; Malins, A.; Yoshimura, Kazuya; Kurikami, Hiroshi; Kitamura, Akihiro; Iijima, Kazuki; Hosomi, Masaaki*
Journal of Environmental Radioactivity, 184-185, p.53 - 62, 2018/04
A study is presented on the applicability of the distribution coefficient () absorption/desorption model to simulate dissolved Cs concentrations in Fukushima river water. The simulation results were in good agreement with the observations on water and suspended sediment fluxes, and on particulate bound Cs concentrations under both ambient and high flow conditions. By contrast the measured concentrations of dissolved Cs in the river water were much harder to reproduce with the simulations. By tuning the values for large particles, it was possible to reproduce the mean dissolved Cs concentrations during base flow periods (observation: 0.32 Bq/L, simulation: 0.36 Bq/L). However neither the seasonal variability in the base flow dissolved Cs concentrations (0.14-0.53 Bq/L), nor the peaks in concentration that occurred during storms (0.18-0.88 Bq/L, mean: 0.55 Bq/L), could be reproduced with realistic simulation parameters.
Kitamura, Akihiro; Imaizumi, Yoshitaka*; Yamaguchi, Masaaki; Yui, Mikazu; Suzuki, Noriyuki*; Hayashi, Seiji*
Kankyo Hoshano Josen Gakkai-Shi, 2(3), p.185 - 192, 2014/09
Annual discharge rates of radioactive cesium through selected rivers due to the Fukushima Dai-ichi Nuclear Power Plant accident were simulated by two different watershed models. One is the Soil and Cesium Transport, SACT, model which was developed by Japan Atomic Energy Agency and the other one is the Grid-Catchment Integrated Modeling System, G-CIEMS, which was developed by National Institute of Environmental Studies. We choose the Abukuma, the Ukedo, and the Niida rivers for the present study. Comparative results showed that while components and assumptions adopted in two models differ, both methods predicted the same order of magnitude estimates.
Asayama, Tai; Miyagawa, Takayuki*; Dozaki, Koji*; Kamishima, Yoshio*; Hayashi, Masaaki*; Machida, Hideo*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
This paper is the first one of the series of four papers that describe ongoing activities in the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) on the elaboration of the System Based Code (SBC) concept. A brief introduction to the SBC concept is followed by the technical features of structural evaluation methodologies that are based on the SBC concept. Also described is the ongoing collaboration of JSME and ASME at the Joint Task Group for System Based Code established in the ASME Boiler and Pressure Vessel Code Committee which is developing alternative rules for ASME B&PV Code Section XI Division 3, inservice inspection requirements for liquid metal reactor components.
Machida, Hideo*; Asayama, Tai; Watanabe, Taigo*; Hojo, Kiminobu*; Hayashi, Masaaki*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
In the System Based Code (SBC), a reliability target is defined according to the importance of components (risk and/or failure probability), and grade of material, design, manufacture and maintenance are chosen to satisfy the reliability target. Therefore, reliability evaluation of components plays the important role of the concept of SBC. Until now, the LRFD methods were developed for burst due to internal pressure, plastic collapse due to membrane and bending stress, fatigue, limit load assessment of flawed pipe, and buckling of thin wall cylinder. This paper describes the action plans of development of the reliability assessment methods and an examination results up to date.
Omichi, Masaaki*; Asano, Atsushi*; Tsukuda, Satoshi*; Takano, Katsuyoshi*; Sugimoto, Masaki; Saeki, Akinori*; Sakamaki, Daisuke*; Onoda, Akira*; Hayashi, Takashi*; Seki, Shu*
Nature Communications (Internet), 5, p.3718_1 - 3718_8, 2014/04
Protein nanowires exhibiting specific biological activities hold promise for interacting with living cells and controlling and predicting biological responses such as apoptosis, endocytosis and cell adhesion. Here we report the result of the interaction of a single high-energy charged particle with protein molecules. Degradation of the human serum albumin nanowires was examined using trypsin. The biotinylated human serum albumin nanowires bound avidin, demonstrating the high affinity of the nanowires. Human serum albumin-avidin hybrid nanowires were also fabricated from a solid state mixture and exhibited good mechanical strength. The biotinylated human serum albumin nanowires can be transformed into nanowires exhibiting a biological function such as avidin-biotinyl interactions and peroxidase activity. The present technique is a versatile platform for functionalizing the surface of any protein molecule with an extremely large surface area.
Matsubara, Masaaki*; Hayashi, Tatsuya*; Suzuki, Masato*; Shiraishi, Taisuke*; Sakamoto, Kenji*; Wakai, Takashi
Nihon Kikai Gakkai M&M 2011 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), 2 Pages, 2011/07
In this study, we investigated the influence of two or more flaws on the collapse strength of austenitic stainless steel piping to improve the safety of a nuclear power plant. The multiple flaws may be more often than single flaw. Therefore, it is important to evaluate the collapse strength of the piping with multiple flaws. Piping with a single flaw and multiple flaws were used in the examination. Collapse strength of the piping is verified under the combination condition of tension and bending. It aims to model from multiple flaws to the single flaw for simplification of the. As a result, (1) It is possible to evaluate the collapse strength by identifying the multiple flaws as the single flaw conservatively. (2) On each load condition, the parallel notches on piping can be accounted to be the single notch.
Takamura, Shuichi*; Kado, Shinichiro*; Fujii, Takashi*; Fujiyama, Hiroshi*; Takabe, Hideaki*; Adachi, Kazuo*; Morimiya, Osamu*; Fujimori, Naoji*; Watanabe, Takayuki*; Hayashi, Yasuaki*; et al.
Kara Zukai, Purazuma Enerugi No Subete, P. 164, 2007/03
no abstracts in English
Ninomiya, Hiromasa; Akiba, Masato; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hayashi, Nobuhiko; Hosogane, Nobuyuki; Ikeda, Yoshitaka; Inoue, Nobuyuki; et al.
Journal of the Korean Physical Society, 49, p.S428 - S432, 2006/12
To contribute DEMO and ITER, the design to modify the present JT-60U into superconducting coil machine, named National Centralized Tokamak (NCT), is being progressed under nationwide collaborations in Japan. Mission, design and strategy of this NCT program is summarized.
Kikuchi, Mitsuru; Tamai, Hiroshi; Matsukawa, Makoto; Fujita, Takaaki; Takase, Yuichi*; Sakurai, Shinji; Kizu, Kaname; Tsuchiya, Katsuhiko; Kurita, Genichi; Morioka, Atsuhiko; et al.
Nuclear Fusion, 46(3), p.S29 - S38, 2006/03
The National Centralized Tokamak (NCT) facility program is a domestic research program for advanced tokamak research to succeed JT-60U incorporating Japanese university accomplishments. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility and mobility is pursued in aspect ratio and shape controllability, feedback control of resistive wall modes, wide current and pressure profile control capability for the demonstration of the high-b steady state.
Tsuchiya, Katsuhiko; Akiba, Masato; Azechi, Hiroshi*; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hashizume, Hidetoshi*; Hayashi, Nobuhiko; Horiike, Hiroshi*; et al.
Fusion Engineering and Design, 81(8-14), p.1599 - 1605, 2006/02
no abstracts in English
Tamai, Hiroshi; Akiba, Masato; Azechi, Hiroshi*; Fujita, Takaaki; Hamamatsu, Kiyotaka; Hashizume, Hidetoshi*; Hayashi, Nobuhiko; Horiike, Hiroshi*; Hosogane, Nobuyuki; Ichimura, Makoto*; et al.
Nuclear Fusion, 45(12), p.1676 - 1683, 2005/12
Design studies are shown on the National Centralized Tokamak facility. The machine design is carried out to investigate the capability for the flexibility in aspect ratio and shape controllability for the demonstration of the high-beta steady state operation with nation-wide collaboration, in parallel with ITER towards DEMO. Two designs are proposed and assessed with respect to the physics requirements such as confinement, stability, current drive, divertor, and energetic particle confinement. The operation range in the aspect ratio and the plasma shape is widely enhanced in consistent with the sufficient divertor pumping. Evaluations of the plasma performance towards the determination of machine design are presented.
Tanaka, Masaaki; Kawashima, Shigeyo*; Igarashi, Minoru; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki
JNC TN9400 2003-117, 65 Pages, 2004/03
Temperature fluctuation due to mixing of hot and cold fluids gives thermal fatigue to the structure (thermal striping phenomena).Investigation of this phenomenon is significant for the safety of a fast breeder reactor, which uses liquid metal as a coolant. In Japan Nuclear Cycle Development Institute, experiments and numerical analyses have been carried out to understand this phenomenon and also to construct the evaluation rule, which can be applied to the design. A water experiment of fluid mixing in T-pipe with long cycle fluctuation (WATLON),which notices thermal striping phenomena in the T-pipe junction, is performed to investigate the key factor of mixing phenomena by reason of long cycle fluctuation observed in a plant. By the former visualization test, it was showed that the flow pattern of branch pipe jet could be classified into (A) impinging jet, (B) deflecting jet (C) re-attachment jet and (D) wall jet according to the inflow condition. It was confirmed that the each jet pattern could be predicted by the momentum ratio of the each piping fluid. In this study, a thermo-chromic liquid crystal sheet was put on the inner wall in the main pipe, and temperature field on the wall surface was visualized. We established a new method to convert the color image data to temperature data. And measurement uncertainty of this method was evaluated + and - about 2.0 [deg-C], using by the typical picture in the temperature calibration test. From the temperature fluctuation visualization test by liquid crystal sheet, the cold spot was formed in just downstream region from the outlet of the branch pipe in the cases of the wall jet and impinging jet. Since this cold spot moved in time, high value of temperature fluctuation intensity was shown around the cold spot. And the validity of this method was shown from the comparison of the thermocouple data installed in a wall surface with the temperature conversion result.
Tamai, Hiroshi; Matsukawa, Makoto; Kurita, Genichi; Hayashi, Nobuhiko; Urata, Kazuhiro*; Miura, Yushi; Kizu, Kaname; Tsuchiya, Katsuhiko; Morioka, Atsuhiko; Kudo, Yusuke; et al.
Plasma Science and Technology, 6(1), p.2141 - 2150, 2004/02
The dominant issue for the the modification program of JT-60 (JT-60SC) is to demonstrate the steady state reactor relevant plasma operation. Physics design on plasma parameters, operation scenarios, and the plasma control method are investigated for the achievement of high-. Engineering design and the R&D on the superconducting magnet coils, radiation shield, and vacuum vessel are performed. Recent progress in such physics and technology developments is presented.
Igarashi, Minoru; Tanaka, Masaaki; Kimura, Nobuyuki; Nakane, Shigeru*; Kawashima, Shigeyo*; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki
JNC TN9400 2003-092, 100 Pages, 2003/11
A water experiment for thermal hydraulics in a mixing tee was performed to investigate thermal striping phenomena. Measurement of flow velocity using particle image velocimetry and temperature measurement were carried out. Normalized power spectrum density of temperature fluctuation had same profile, when the momentum ratio of the main and branch pips is the same. From the velocity measurement test, when the momentum ratio is the same, flow pattern at mixing region shows the alomost same tendency. Temperature transfer characteristics from fluid to structure can be estimated by a constant heat transfer coefficient in time.
Miya, Naoyuki; Kikuchi, Mitsuru; Ushigusa, Kenkichi; ; Nagashima, Keisuke; Neyatani, Yuzuru; Tobita, Kenji; ; Masaki, Kei; Kaminaga, Atsushi; et al.
JAERI-Research 98-012, 222 Pages, 1998/03
no abstracts in English
; Ushigusa, Kenkichi; Kikuchi, Mitsuru; Nagashima, Keisuke; Neyatani, Yuzuru; Miya, Naoyuki; ; Takahashi, Yoshikazu; Hayashi, Takumi; Kuriyama, Masaaki; et al.
Proc. of 17th IEEE/NPSS Symposium Fusion Engineering (SOFE'97), 1, p.233 - 236, 1998/00
no abstracts in English
Kikuchi, Mitsuru; ; ; Miya, Naoyuki; Ushigusa, Kenkichi; Nagashima, Keisuke; Aoyagi, Tetsuo; ; Neyatani, Yuzuru; Naito, Osamu; et al.
JAERI-Research 97-026, 70 Pages, 1997/03
no abstracts in English
Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.
PNC TN9410 96-279, 51 Pages, 1996/08
Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.