Tanaka, Yasunori*; Smirnov, R. D.*; Pigarov, A. Y.*; Takenaga, Hidenobu; Asakura, Nobuyuki; Uesugi, Yoshihiko*; Ono, Noriyasu*
Journal of Nuclear Materials, 415(Suppl.1), p.S1106 - S1110, 2011/08
Dust can be an important contributor to impurity contamination of the core and scrape-off-layer (SOL) plasmas in tokamak fusion devices. This is the first report about investigation of transport of carbon dust particles in case of JT-60U tokamak using the dust transport code (DUSTT). The DUSTT code takes into account various plasma-dust interaction processes. In the present report, background plasma parameters in JT-60U such as densities and temperatures of electron, ion and neutral atom were computed with the UEDGE code. Three dimensional trajectories, temperature evolution, and radius variation of dust particles launched from different positions at the inner divertor, the outer divertor and the dome structure in JT-60U tokamak were simulated numerically to study dynamics and transport of dust particles there. As a result, the lifetime of dust particles is dependent mainly on the ion density in trajectories in final term.
Baba, Hiromi*; Onizuka, Yoshihiko*; Nakao, Minoru*; Fukahori, Mai*; Sato, Tatsuhiko; Sakurai, Yoshinori*; Tanaka, Hiroki*; Endo, Satoru*
Radiation Protection Dosimetry, 143(2-4), p.528 - 532, 2011/02
The PHITS simulation were performed to reproduce the geometrical setup of an experiment that measured the microdosimetric energy distributions at the Kyoto University Reactor (KUR) where two types of tissue equivalent proportional counters (TEPC) were used, one with A-150 wall alone and another with a 50 ppm boron loaded A-150 wall. It was found that the PHITS code is a useful tool for the simulation of the energy deposited in tissue in BNCT based on the comparisons with experimental results.
Ida, Katsumi*; Sakamoto, Yoshiteru; Yoshinuma, Mikiro*; Takenaga, Hidenobu; Nagaoka, Kenichi*; Hayashi, Nobuhiko; Oyama, Naoyuki; Osakabe, Masaki*; Yokoyama, Masayuki*; Funaba, Hisamichi*; et al.
Nuclear Fusion, 49(9), p.095024_1 - 095024_9, 2009/09
Dynamics of ion internal transport barrier (ITB) formation and impurity transport both in the Large Helical Device (LHD) heliotron and JT-60U tokamak are described. Significant differences between heliotron and tokamak plasmas are observed. The location of the ITB moves outward during the ITB formation regardless of the sign of magnetic shear in JT-60U and the ITB becomes more localized in the plasma with negative magnetic shear. In LHD, the low Te/Ti ratio ( 1) of the target plasma for the high power heating is found to be necessary condition to achieve the ITB plasma and the ITB location tends to expand outward or inward depending on the condition of the target plasmas. Associated with the formation of ITB, the carbon density tends to be peaked due to inward convection in JT-60U, while the carbon density becomes hollow due to outward convection in LHD. The outward convection observed in LHD contradicts the prediction by neoclassical theory.
Tanaka, Hirohiko*; Ono, Noriyasu*; Asakura, Nobuyuki; Tsuji, Yoshiyuki*; Kawashima, Hisato; Takamura, Shuichi*; Uesugi, Yoshihiko*; JT-60U Team
Nuclear Fusion, 49(6), p.065017_1 - 065017_7, 2009/06
Comparison between fluctuation characteristics at high-field-side (HFS) and low-field-side (LFS) scrape off layers (SOLs) has been made, for the first time, in the L mode plasma of the JT-60U tokamak using reciprocating Langumuir probes. Statistical analysis based on probability distribution function (PDF) was employed to describe intermittent (non-diffusion) transport in SOL plasma fluctuations. The positive bursty events appeared most frequently at LFS midplane associated with blobby plasma transport, then the PDF is strongly skewed positively, while the PDF in HFS SOL is close to Gaussian distribution. Conditional averaging analysis of the positive bursty events at LFS midplane indicates the intermittent feature with a rapid increase and a slow decay is similar to that of plasma blobs theoretically predicted. Statistical self-similarity was also investigated with Fourier power spectrum and statistics of waiting-time and duration-time of the fluctuation.
Reyes-Borja, W. O.*; Sotomayor, I.*; Garzn, I.*; Vera, D.*; Cedeo, M.*; Tanaka, Atsushi; Hase, Yoshihiro; Sekozawa, Yoshihiko*; Sugaya, Sumiko*; Gemma, Hiroshi*
JAEA-Review 2007-060, JAEA Takasaki Annual Report 2006, P. 89, 2008/03
Endo, Satoru*; Tanaka, Kenichi*; Takada, Masashi*; Onizuka, Yoshihiko*; Miyahara, Nobuyuki*; Sato, Tatsuhiko; Ishikawa, Masayoshi*; Maeda, Naoko*; Hayabuchi, Naofumi*; Shizuma, Kiyoshi*; et al.
Medical Physics, 34(9), p.3571 - 3578, 2007/09
Absorbed doses from main charged particle beams and charged-particle fragments have been measured with high accuracy for particle therapy but there are few reports for doses from neutron components produced as fragments. This study describes measurements on neutron dose produced by carbon beam, microdosimetric distributions of secondary neutrons produced by 290 MeV/nucleon carbon beams have been measured by using a tissue equivalent proportional counter (TEPC) at the Heavy Ion Medical Accelerator in Chiba (HIMAC) at the National Institute of Radiological Sciences (NIRS). The ratios of neutrons to charged particle fragments dominated to be 11 to 89 % in the absorbed doses at the side and below the faces of the acrylic phantom (300 mm height 300 mm width 253 mm thickness).
Reyes-Borja, W. O.*; Sotomayor, I.*; Garzn, I.*; Vera, D.*; Cedeo, M.*; Castillo, B.*; Tanaka, Atsushi; Hase, Yoshihiro; Sekozawa, Yoshihiko*; Sugaya, Sumiko*; et al.
Plant Biotechnology, 24(3), p.349 - 353, 2007/06
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
Okane, Tetsuo; Okamoto, Jun*; Mamiya, Kazutoshi*; Fujimori, Shinichi; Takeda, Yukiharu; Saito, Yuji; Muramatsu, Yasuji*; Fujimori, Atsushi*; Haga, Yoshinori; Yamamoto, Etsuji; et al.
Journal of the Physical Society of Japan, 75(2), p.024704_1 - 024704_5, 2006/02
no abstracts in English
Sukekawa, Masayuki*; Isobe, Nobuhiro*; Shibamoto, Hiroshi; Tanaka, Yoshihiko*; Kasahara, Naoto
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 5 Pages, 2006/00
For expansion of non-creep design area and simplification of design procedures, a rational identification method of creep design area by negligible creep (NC) curves was studied. NC curves of six kinds of stainless and ferrite steels for fast reactors were determined at 1.5Sm (Sm: design stress intensity). These NC curves are based on domestic material data. NC curves provide the relation between temperature and time that does not induce meaningful creep strain under the constant primary stress. As for 316FR steel, which is used for reactor vessel in Japanese fast reactor, non-creep design area is identified with comparing the highest temperature and 425C (constant upper limit for austenite stainless steal) by existing Japanese Guides. However, this temperature limit can be enhanced by NC curve concept when operating (thermal transient) time is long. NC curves under higher primary stress, and the curves under secondary stress were also studied. However, at the present stage, NC curves for stress level 1.5Sm were adopted to identify creep design area. The concept of NC curve was introduced into the interim FDS (fast reactor design standard for commercialized fast reactors in Japan) to simplify the creep design of fast reactor systems. Utilizing these curves, design becomes easier for components which are employed at comparatively lower temperature under normal condition and short holding time at high temperature.
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.
Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai; Inoue, Kazuhiko*; Shibamoto, Hiroshi*; Tanaka, Yoshihiko*
JNC TY9400 2004-025, 984 Pages, 2004/08
Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company(JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R&D activities for the feasibility studies on commerdalized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study, This report contains the results of the research on the structural design technology. The research scope was identified as (1) FDS(FBR Design Standard), (2) Standardization of new material, and (3)System Based Code for Integrity, and the results of this year's studies are summarized as follows. (1)FDS (FBR Design Standard) * As for failure criteria, ratcheting-fatigue tests were continued. Applicability of rational settling method on creep design regime was evaluated and evaluation method of primary stress was studied. * As for a guideline on inelastic analysis for design, development of conservative detail modle (CRIEPI model for design) is underway. Loading history effect was evaluated through analysis. Conservative evaluation method of creep-fatigue damage coped with inelastic analysis was also developed. Aiming for verification of the guidline, structure model test simulated sodium surface level of reactor vessel is continuing. Policy and items of the guideline were studied. * As for a guideline on thermal loads modeling for design, provisions of the guideline on rational settling method of thermal striping loads were discussed. Screening method to grasp severe thermal load and parts in higher stress was developed. (2)Standardization of new material * As for candidate 12-chromium stainless steel (added tungsten, non-added tungsten), that is expected to improve strength of components of commercialized fast reactor, short and medium-term strength tests (including long-term aged test piece), ob
kasahara, Naoto; Ando, Masanori; Ito, Kei; Tanaka, Yoshihiko; Shibamoto, Hiroshi; Inoue, Kazuhiko
ASME PVP-Vol.472, p.25-32, p.25 - 32, 2004/07
For the realization of safe and economical fast reactor (FR) plants, the Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on a research project titled "Feasibility Study on Commercialized FR Cycle Systems. To certify the design concepts and validate their structural integrity,the research and development of the "Fast Reactor Structural Design Standard (FDS)" is recognized as begin an essential theme.
Tanaka, Yoshihiko; Shibamoto, Hiroshi; Inoue, Kazuhiko; kasahara, Naoto; Ando, Masanori; Ito, Kei
ASME PVP-Vol.472, p.53-60, p.53 - 60, 2004/07
The guideline on inelastic analysis for design, one of the key items of Fast Reactor Design Standard(FDS), is being developed.The basic policies of this guideline are as follows:(a) to emphasis conservative analysis output rather than nominal value representing actual behavior, (b) to clarify the applicable area for assurance of conservative results. With such concepts, it would be possible that the guideline provides useful explanations on the manner of analysis and estimation in the form of concrete examples of design as well as general rules (somehow vague). As the first step of the guideline development, the following five issues to be solved were extracted:1) applicable area, 2) selection of constitutive equation, 3) modeling method of the load history, 4) ratchet strain and creep fatigue damage evaluation methods by inelastic analysis and 5) example design problems to check users' analysis quality and to complement the general rules. In parallel, inelastic analyses with the promising constitutive equations were applied by way of trial to obtain rough presumption on their effects on structural design of the components. As a result,all inelastic analyses provided smaller cumulative strains and equivalent strain ranges than the existing design method based on elastic analysis,suggesting advantage of introducing them into actual design.
Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.
JNC TN9400 2004-035, 2071 Pages, 2004/06
The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.
Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai; Sagayama, Yutaka*; Inoue, Kazuhiko*; Shibamoto, Hiroshi*; Tanaka, Yoshihiko*
JNC TY9400 2003-001, 644 Pages, 2003/05
Tanaka, Yoshihiko; Kasahara, Naoto
JNC TN9400 2003-037, 95 Pages, 2003/05
The advanced loop-type reactor system, one of the promising concepts in the Feasibility study of the FBR Cycle, adopts many innovative ideas to meet the challenging requirements for safety and economy. As a results, it seems that the structures of the reactor system would be subjected to severer loads than the predecessors. 0ne of the countermeasures to them is the design by inelastic analysis. In the past, many studies showed that structural design by inelastic analysis is much more reasonable than one by conservative elastic analysis. However, inelastic analysis has hardly been adopted in nuclear design so far. One of the reasons is that inelastic analysis has loading history effect, that is, the analysis result would differ depending on the order of loads. It seems to be difficult to find the general solution for the loading history effect. Consequently, inelastic analysis output from the four deferent thermal load histories which consists of the thermal load cycle including the severest cold shock ("C")and the one including the severest hot shock ("H") were compared with each other. From this comparison, it was revealed that the thermal load history with evenly distributed "H"s among "C" s tend to give the most conservative damage estimation derived from inelastic analysis output. Therefore, such thermal load history pattern is proposed for the structural design by inelastic analysis.
Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai; Sagayama, Yutaka*; Dozaki, Koji*; Shibamoto, Hiroshi*; Tanaka, Yoshihiko*
JNC TY9400 2002-025, 889 Pages, 2003/01