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Sakurai, Junya*; Torigata, Keisuke*; Matsunaga, Manabu*; Takanashi, Naoto*; Hibino, Shinya*; Kizu, Kenichi*; Morita, Akira*; Inomoto, Masahiro*; Shimohata, Nobuaki*; Toyota, Kodai; et al.
Tetsu To Hagane, 111(5), p.246 - 262, 2025/04
Kudo, Yusuke; Sawai, Tomotsugu; Sakurai, Shinji; Masaki, Kei; Suzuki, Yutaka; Sasajima, Tadayuki; Hayashi, Takao; Takahashi, Ryukichi*; Honda, Masao; Jitsukawa, Shiro; et al.
Journal of the Korean Physical Society, 49(96), p.S297 - S301, 2006/12
Installation of ferritic steel tiles was proposed in JT-60U to reduce the toroidal magnetic field ripple and to improve the fast ion loss, which degrades heating efficiency and increases heat load on plasma facing component under large volume plasma operations. We selected a 8Cr-2W-0.2V ferritic steel with the cost-effectiveness, in which concentration limits of activation elements in F82H were relaxed because of the less number of neutron generations from deuterium operations on JT-60U. The fabricated ferritic steel has clear tempered martensitic microstructure, and sufficient magnetic and mechanical properties. The saturated magnetization was estimated to 1.7 Tesla at 573 K, lower than expected, but effectiveness in JT-60U was confirmed by numerical analyses. To research the effect of material conditions, such as microstructure and heat treatment, on saturated magnetization of the ferritic steel based on 8-9Cr is important for the future fusion reactors which will be planned to install the ferritic steel as the in-vessel components.
Suzuki, Satoshi; Ezato, Koichiro; Hirose, Takanori; Sato, Kazuyoshi; Yoshida, Hajime; Enoeda, Mikio; Akiba, Masato
Fusion Engineering and Design, 81(1-7), p.93 - 103, 2006/02
Times Cited Count：13 Percentile：64.62(Nuclear Science & Technology)This paper presents an R&D activity on the plasma facing components (PFCs), such as first wall and divertor, for the fusion power plant. The PFCs of the power plant will be subjected to heavy neutron irradiation and high heat/particle flux from plasma during the continuous operation. In the present design of the PFCs, the candidate structural material is a reduced activation ferritic-martensitic steel, F82H, from the viewpoints of low activation and high robustness against neutron irradiation, and the candidate armor material is tungsten from the low sputtering yield and low tritium retention points of view. To realize the PFCs using such materials, JAERI has bee extensively conducting R&Ds on; (1) high performance cooling tube, (2) tungsten armor materials, (3) selection of a bonding technique for F82H and tungsten materials and (4) evaluation of structural integrity. Recent achievements on these R&Ds are presented.
Yamamoto, Masahiro*; Shibata, Takatoshi; Tsuzuki, Kazuhiro; Sato, Masayasu; Kimura, Haruyuki; Okano, Fuminori; Kawashima, Hisato; Suzuki, Sadaaki; Shinohara, Koji; JFT-2M Group; et al.
Fusion Science and Technology, 49(2), p.241 - 248, 2006/02
Times Cited Count：2 Percentile：16.71(Nuclear Science & Technology)JFT-2M has been modified three times in the Advanced Material Tokamak Experiment (AMTEX) program to investigate compatibility of the low activation ferritic steel F82H with tokamak plasmas as a structural material for future reactors. The ferritic steel plate/wall was installed inside and/or outside of the vacuum vessel to reduce the ripple of toroidal magnetic field step by step through three modifications. This paper focuses on engineering aspects in these modifications; electromagnetic analysis to find a suitable way for fixing these plates, installation procedure to keep small tolerance, a three-dimensional magnetic field measurement device used to obtain information of the actual shape of the vacuum vessel used as a installation standard surface. To keep a good surface condition of the ferritic steel plate/wall that rusts easily, careful treatment was executed before the installation. To reduce oxygen impurities further, a boronization system with tri-methyl boron, which is safe and easy to operate, was developed.
Yutani, Toshiaki*; Nakamura, Hiroo; Sugimoto, Masayoshi
JAERI-Tech 2005-036, 10 Pages, 2005/06
JAERI-Tech-2005-036.pdf:2.06MB
In the high flux region of the International Fusion Materials Irradiation Facility (IFMIF), the neutron irradiation damage for iron-based alloys will exceed 20 dpa/ year. An accurate specimen temperature measurement under a large amount of nuclear heating is a key issue but the change of heat transfer of gap between irradiation specimens and specimen holder during irradiation test is inevitable, if gap is filled with an inert gas and temperature is monitored by a thermocouple buried in the specimen holder. A solution to make heat transfer predictable is to fill the gap with a liquid metal (sodium or sodium-potassium alloy). An issue of compatibility between Reduced Activation Ferritic/Martensitic steels and the liquid metalsis addressed in this paper, and some recommendations for designing irradiation rig are presented, such as a purification control before filling liquid metals, or a careful selection of material of rig to avoid carbon mass transfer.
Yoshida, Hajime; Kosaku, Yasuo*; Enoeda, Mikio; Abe, Tetsuya; Akiba, Masato
JAERI-Research 2005-003, 13 Pages, 2005/03
JAERI-Research-2005-003.pdf:3.33MB
Hydrogen permeation fluxes of the reduced activation ferritic steel F82H were quantitatively measured by a newly proposed method, vacuum thermo-balance method, for a precise estimation of tritium leakage in a fusion reactor. We prepared sample capsules made of F82H, which enclosed hydrogen gas. The hydrogen in the capsules permeated through the capsule wall, and subsequently desorbed from the capsule surface during isothermal heating. The vacuum thermo-balance method allows simultaneous measurement of the hydrogen permeation flux by two independent methods, namely, the net weight reduction of the sample capsule and exhaust gas analysis. Thus the simultaneous measurements by two independent methods increase the reliability of the permeability measurement. The ratio of the hydrogen permeation fluxes obtained by the net weight reduction to that measured by the exhaust gas analysis was in the range from 1/4 to 1/1 in this experiment. It has been demonstrated that the vacuum thermo-balance method is effective for the measurement of hydrogen permeation rate of F82H.
Tanigawa, Hiroyasu; Sakasegawa, Hideo*; Klueh, R. L.*
Materials Transactions, 46(3), p.469 - 474, 2005/03
Times Cited Count：22 Percentile：74.85(Materials Science, Multidisciplinary)The effects of irradiation on precipitation of reduced-activation ferritic/martensitic steels (RAFs) were investigated, and its impacts on the Charpy impact properties and tensile properties were discussed. It was previously reported that RAFs (F82H-IEA and its heat treatment variant, ORNL9Cr-2WVTa, JLF-1 and 2%Ni doped F82H) shows variety of changes on its ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573K up to 5dpa. These differences could not be interpreted as an effect of irradiation hardening caused by dislocation loop formation. The precipitation behavior of the irradiated steels was examined by weight analysis, X-ray diffraction analysis and chemical analysis on extraction residues. These analyses suggested that irradiation caused (1) the increase of the amount of precipitates (mainly M
C
), (2) increase of Cr and decrease of W contained in precipitates, (3) disappearance of MX (TaC) in ORNL9Cr and JLF-1.
Miura, Yukitoshi; Hoshino, Katsumichi; Kusama, Yoshinori
Purazuma, Kaku Yugo Gakkai-Shi, 80(8), p.653 - 661, 2004/08
A series of experimental program on the JAERI Fusion Torus-2M (JFT-2M) was completed in March, 2004. In the experimental operation for 21 years since the first plasma on April 27, 1983, many significant results leading the fusion energy research and plasma physics have been produced in researches on high confinement mode (H-mode), heating and current drive, advanced plasma control, compatibility of low activation ferritic steel with improved confinement mode, etc. Among these results, some important results are presented.
Matsuhiro, Kenjiro; Ando, Masami; Nakamura, Hiroo; Takeuchi, Hiroshi
JAERI-Research 2004-003, 12 Pages, 2004/03
JAERI-Research-2004-003.pdf:0.85MB
The effect of neutron irradiation damage on tritium permeation through reduced-activation ferritic steel (F82H) at IFMIF target backwall has been estimated. From the results, it has been found that the effective diffusion coefficient of hydrogen in F82H will decrease by 10 % to 20 % under neutron irradiation. Therefore, the amount of tritium permeation for several hundred seconds at the beginning of permeation will be smaller than 80 % to 90 % of that before neutron irradiation. The amount of tritium permeation of F82H at IFMIF target backwall is 1.3x10
g/d (4.7x10
Bq/d). It is 30 times larger than that of 316SS, and is about 8 % of tritium permeation at main loop of IFMIF.
Urata, Kazuhiro*; Shinohara, Koji; Suzuki, Masanobu*; Kamata, Isao*
JAERI-Data/Code 2004-007, 45 Pages, 2004/03
JAERI-Data-Code-2004-007.pdf:5.63MB
As the toroidal magnetic field generated by discrete TF coils involves magnetic field ripple, the fast ion loss is induced to damage vacuum vessel in tokamaks. An idea of ripple compensation using ferromagnetic is proposed. Since low activation ferritic steel have low activation and thermal conduction properties, the ferritic steel is planned to install in tokamak reactors. Installation of ferritic steel plates with toroidal symmetry is effective to compensate ripple, however in the actual devices it is difficult for interference with other components. Besides the first wall shapes are often asymmetric. So it is better to treat toroidal asymmetry to evaluate the ripple induced loss in the actual devices. For the purpose, magnetic field calculation code considering ferritic steel; FEMAG(FErrite generating MAGnetic field)has been speeded up. On the basis of this magnetic field data, OFMC (Orbit Following Monte Carlo) has been upgraded to treat toroidal asymmetry. The use of FEMAG/OFMC, applications to the JFT-2M experiments, and the national centralized tokamak facility are reported.
The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi
JAERI-Tech 2004-012, 62 Pages, 2004/02
JAERI-Tech-2004-012.pdf:16.4MB
At Onagawa Nuclear Power Station Unit-1 of the Tohoku Electric Power co., inc., cracks were confirmed near welded joints of core shroud in 15th periodical inspection. Tohoku Electric Power co., inc. has conducted a material examination with Nippon Nuclear Fuel Development Co., Ltd.. To investigate independently, a JAERI's own evaluation report was provided. The results are as follows; (1) Hardening layer was detected at the depth of about 150-250
m from outer surface of the sample. (2) Corrosion products were observed on inner surface of the cracks and some of them penetrated into grains. (3) Transgranular cracking and intergranular cracking were observed at the region within about 100m and the deeper region more than about 200m in depth from outer surface of the sample, respectively. (4) Distinct chromium depletion was not detected at the grain boundaries. (5) Chemical compositions of the sample corresponded to type 304L stainless steel in Japanese Industrial Standard. From the above, it is concluded that the cracks are stress corrosion cracking.
Ishida, Shinichi; Abe, Katsunori*; Ando, Akira*; Cho, T.*; Fujii, Tsuneyuki; Fujita, Takaaki; Goto, Seiichi*; Hanada, Kazuaki*; Hatayama, Akiyoshi*; Hino, Tomoaki*; et al.
Nuclear Fusion, 43(7), p.606 - 613, 2003/07
Times Cited Count：33 Percentile：68.09(Physics, Fluids & Plasmas)no abstracts in English
Research Committee for Fusion Reactor; Research Committee for Fusion Materials
JAERI-Review 2003-015, 123 Pages, 2003/05
JAERI-Review-2003-015.pdf:24.89MB
no abstracts in English
Urata, Kazuhiro*
JAERI-Data/Code 2003-005, 36 Pages, 2003/03
JAERI-Data-Code-2003-005.pdf:2.96MB
In design of the fusion devises in which ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect and the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of ferritic magnetic field. However iterative calculations by the non-linearity in B-H curve disturbs high-speed calculation. In the strong toroidal magnetic field in the tokamak, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward. Additionally objective ferritic steel geometry is limited to the thin plate and they are installed along the toroidal magnetic field. Taking them into account, high-speed calculation code FEMAG has been developed. In this report, the formalization of FEMAG, how to use FEMAG, and the validity check in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are design studies for JT-60 modification.
Jitsukawa, Shiro; Tamura, Manabu*; Van der Schaaf, B.*; Klueh, R. L.*; Alamo, A.*; Petersen, C.*; Schirra, M.*; Spaetig, P.*; Odette, G. R.*; Tavassoli, A. A.*; et al.
Journal of Nuclear Materials, 307-311(Part1), p.179 - 186, 2002/12
Times Cited Count：171 Percentile：99.28(Materials Science, Multidisciplinary)Reduced activation ferritic/martensitic steel is the primary candidate structural material for ITER Test Blanket Modules and DEMOnstration fusion reactor because of its excellent dimensional stability under irradiation and lower residual activity as compared with the Ni bearing steels such as the austenitic stainless steels. In this paper, microstructural features, tensile, fracture toughness, creep and fatigue properties of a reduced activation martensitic steel F82H (8Cr-2W-0.04Ta-0.1C) are reported before and after irradiation, in addition to the design concept used for development of this alloy. A large number of collaborative test results including those generated under the IEA working group implementing agreements are collected and are used to evaluate the feasibility of use of F82H steel as one of the reference alloys. The effect of metallurgical variables on the irradiation hardening is reviewed and compared with the results obtained from irradiation experiments.
Terada, Yasuaki*; Ochiai, Kentaro; Sato, Satoshi; Wada, Masayuki*; Klix, A.; Yamauchi, Michinori*; Hori, Junichi; Nishitani, Takeo
JAERI-Research 2002-019, 70 Pages, 2002/10
JAERI-Research-2002-019.pdf:8.47MB
D-T neutron irradiation experiments have been carried out with a F82H-containing breeding blanket mock-up of a fusion in order to investigate the activation characteristics of F82H low activation stainless steel. We have measured reaction rates producing 54Mn, 56Mn, 51Cr and 187W in foils of F82H, chromium and tungsten. MCNP calculations were done with evaluated nuclear data from the JENDL-3.2 and the FENDL/E-2.0 files and the results were compared with the measured values. The comparison shows that by using the current data files the reaction rates obtained from the calculations will be overestimated by up to 10-20% for 54Mn, 56Mn and 51Cr, up to 30-40% for 187W, respectively. The calculated values for tungsten are different with the evaluated nuclear data library, which shows that the neutron capture cross sections of tungsten have discrepancy in the resonance region for each nuclear data libraries.
Ninomiya, Hiromasa; Kitsunezaki, Akio; Shimizu, Masatsugu; Kuriyama, Masaaki; JT-60 Team; Kimura, Haruyuki; Kawashima, Hisato; Tsuzuki, Kazuhiro; Sato, Masayasu; Isei, Nobuaki; et al.
Fusion Science and Technology, 42(1), p.7 - 31, 2002/07
Times Cited Count：13 Percentile：24.24(Nuclear Science & Technology)In order to establish scientific basis for the sustainment of highly integrated performance required in the advanced steady-state operation, JT-60U has been optimizing the discharge control scenarios of improved confinement plasmas and expanding the operation regions. Promising results toward the steady-state tokamak were obtained. The detail of such results is reported. JFT-2M has performed advanced and basic research for the development of high performance tokamak plasma as well as the structural material for a fusion reactor. The toroidal ripple reduction with ferritic steel plates outside the vacuum vessel was successfully demonstrated. No adverse effects were observed in the pre-testing on compatibility between ferritic steel plates, covering ~20% of the inside wall of the vacuum vessel, and plasma. The results of TRIAM-1M is also reported.
Research Committee for Fusion Reactor; Research Committee for Fusion Materials
JAERI-Review 2002-008, 79 Pages, 2002/03
JAERI-Review-2002-008.pdf:9.92MB
Joint research committee for fusion reactor and materials was held in Tokyo on July 16, 2001. In the committee, a review of the development programs and the present status on the blanket technology, materials and IFMIF(International Fusion Materials Irradiation Facility) in JAERI and Japanese Universities was reported, and the direction of these R&D was discussed. Moreover, the progress of the collaboration between JAERI and Japanese Universities was discussed. This report consists of the summaries of the presentations and the viewgraphs which were used at the committee.
Takatsu, Hideyuki
Purazuma, Kaku Yugo Gakkai-Shi, 74(5), p.434 - 435, 1998/05
no abstracts in English
; Abe, Mitsushi*; Tadokoro, Takahiro*; Miura, Yukitoshi; Suzuki, Norio; Sato, Masayasu; Sengoku, Seio
Purazuma, Kaku Yugo Gakkai-Shi, 74(3), p.274 - 283, 1998/03
no abstracts in English
