Ando, Masami; Nozawa, Takashi; Hirose, Takanori; Tanigawa, Hiroyasu; Wakai, Eiichi; Stoller, R. E.*; Myers, J.*
Fusion Science and Technology, 68(3), p.648 - 651, 2015/10
Pressurized tubes of F82H and B-doped F82H irradiated at 573 and 673 K up to 6dpa have been measured by a laser profilometer. The irradiation creep strain in F82H irradiated at 573 and 673 K was almost linearly dependent on the effective stress level for stresses below 260 MPa and 170 MPa, respectively. The creep strain of BN-F82H was similar to that of F82H IEA at each effective stress level except 294 MPa at 573 K irradiation. For 673 K irradiation, the creep strain of some BN-F82H tubes was larger than that of F82H tubes. It is suggested that a swelling caused in each BN-F82H because small helium babbles might be produced by a reaction of B(n, ) Li.
Wakai, Eiichi; Ando, Masami; Okubo, Nariaki
Journal of Plasma and Fusion Research SERIES, Vol.11, p.104 - 112, 2015/03
The reduced-activation ferritic/martensitic (RAFM) steels for the fusion DEMO reactor have been developing from around the 1980s. RAFM steels are the first candidate materials for the first wall and blanket structure of fusion DEMO reactors, the target back-plate and the target assembly of IFMIF. In this study, two subjects had been examined and are summarized as below: (1) Effect of initial heat treatment on the microstructures and mechanical properties of RAFM steels, including irradiation damage, is very important to design the fusion DEMO reactors and also control the changes of mechanical properties after the irradiation. (2) Effects of He and H production on the microstructures and mechanical properties of RAFM steels, including irradiation damage, are essential in the evaluation of design of fusion DEMO reactor, and we have to check and evaluate them in Fusion irradiation environment like IFMIF.
Sakasegawa, Hideo; Tanigawa, Hiroyasu; Ando, Masami
Journal of Nuclear Science and Technology, 51(6), p.737 - 743, 2014/06
Oxide-dispersion-strengthened (ODS) steels are attractive materials for the fuel cladding of fast reactors and the first-wall material of fusion blanket. High-chromium ferritic ODS steels have better corrosion-resistance properties, but they have poor material workability and anisotropy, making their practical application difficult. In contrast, low-chromium ferritic/martensitic ODS steels have better workability and their anisotropy can be reduced through martensitic transformation. However, their corrosion-resistance properties are poor, compared to high-chromium ferrtic ODS steels. In this work, we developed a corrosion-resistant coating technique for 8Cr ferritic/martensitic ODS steel. The ODS steel was coated with 304 or 430 stainless steel by changing the canning material from mild steel to stainless steel in the conventional material processing procedure and using it as a coating material.
Ando, Masami; Nozawa, Takashi; Hirose, Takanori; Tanigawa, Hiroyasu
Purazuma, Kaku Yugo Gakkai-Shi, 90(1), p.64 - 67, 2014/01
Reduced activation ferritic/martensitic steel (RAFM) is a candidate for the material of DEMO blanket structure. The irradiation creep behavior of F82H and JLF-1 steel has been measured at 300, 400 and 500C up to 5 dpa using helium-pressurized creep tubes irradiated in HFIR. These tubes were pressurized with helium to hoop stress levels of 0400 MPa at the irradiation temperature. The results for F82H and JLF-1 with a 400 MPa hoop stress detected small creep strains ( 0.25%) after irradiation at 300C. Irradiation creep rate (creep strain/dose) was tendency to be a similar behavior for high-dose irradiated RAFM specimens in FFTF. In this paper, a procedure of irradiation creep test & evaluation was also summarized.
Hirose, Takanori; Sokolov, M. A.*; Ando, Masami; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Stoller, R. E.*; Odette, G. R.*
Journal of Nuclear Materials, 442(1-3), p.S557 - S561, 2013/11
Hirose, Takanori; Okubo, Nariaki; Tanigawa, Hiroyasu; Ando, Masami; Sokolov, M. A.*; Stoller, R. E.*; Odette, G. R.*
Journal of Nuclear Materials, 417(1-3), p.108 - 111, 2011/10
This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels irradiated at 573 K. The materials used in this research were F82H-IEA and its modified steels. Post irradiation mechanical tests revealed that irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated proof stress is less than 1 GPa. The deterioration of total elongation was also saturated by 9 dpa. Irradiation response of F82H-mod3, which is stable to temperature instability during material production and HIP treatment, was very similar to that of F82H-IEA, and negative impacts of extra tantalum was not observed. Therefore it can be an attractive option for the structural materials for blanket components manufactured by HIP.
Terashima, Takaya*; Motokawa, Ryuhei; Koizumi, Satoshi*; Sawamoto, Mitsuo*; Kamigaito, Masami*; Ando, Tsuyoshi*; Hashimoto, Takeji*
Macromolecules, 43(19), p.8218 - 8232, 2010/10
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.
Hirose, Takanori; Ando, Masami; Ogiwara, Hiroyuki*; Tanigawa, Hiroyasu; Enoeda, Mikio; Akiba, Masato
Fusion Engineering and Design, 85(5), p.809 - 812, 2010/08
In this work, the interfacial properties of Be-reduced activation ferritic/martensitic steel (RAFMs) joints were investigated for the first wall of an ITER test blanket module (TBM). The joints were produced by the solid state hot isostatic pressing (HIP) method. Chromium (Cr) was used as a diffusion barrier with a thickness of 1 micron or 10 microns, formed by plasma vapor deposition on the Be surface. The HIPping was conducted at 1023 K and 1233 K. The temperatures are standard normalizing and tempering temperatures of F82H. EPMA showed the Cr layer effectively worked as a diffusion barrier at 1023 K. However, for the F82H/Be interface which underwent HIP at 1233 K followed by tempering a Be rich layer was formed. Bend tests revealed that a thin Cr layer and low temperature HIP is preferable.
Tanigawa, Hiroyasu; Nozawa, Takashi; Ando, Masami
JAEA-Review 2009-053, 89 Pages, 2010/03
R&D on DEMO engineering was started from 2007 as the main activity of a DEMO design and R&D coordination center activity which was one of the international nuclear fusion energy research center (IFERC) activity of "Broader Approach" (BA) activity. In the DEMO engineering R&D activity, five "generic" R&D items were selected, which attract both Japan and EU interest and not specific to a DEMO design, i.e., (1) R&D on SiC/SiC Composites: (2) R&D on Tritium Technology: (3) R&D on Materials Engineering for DEMO Blanket: (R&D on reduced activation ferritic/martensitic steel: RAFM) (4) R&D on Advanced Neutron Multiplier for DEMO Blanket: (5) R&D on Advanced Tritium Breeders for DEMO Blanket. In first phase (2007 -2009), some part of preliminary R&D, which was defined in procurement arrangement (PA) between Japan and EU, is conducted by JAEA university collaboration. This report summarized the results from joint research activities about R&D on SiC/SiC composite and RAFM in 2008.
Jitsukawa, Shiro; Suzuki, Kazuhiko; Okubo, Nariaki; Ando, Masami; Shiba, Kiyoyuki
Nuclear Fusion, 49(11), p.115006_1 - 115006_8, 2009/11
Irradiation often causes hardening and reduction of elongation as well as toughness degradation to a considerable degree. Data, however, indicate that these changes remain in manageable ranges for ITER-TBM application. Moreover, the saturation tendency of the changes with neutron dose suggests that some of the reduced activation martensitic steels are feasible even for future DEMO applications. It is also stressed that the development of a design methodology that is compatible with the large irradiation induced changes is essential to enable these applications. Modeling activities for the macroscopic mechanical response are expected to play key roles in design methodology development. Macroscopic models of plasticity (a constitutive equation) and cyclic softening behavior after irradiation are discussed. Significance of models to estimate microstructural changes during irradiation and beneficial effects of the heat treatment for irradiation performance are also introduced.
Nagashima, Nobuo*; Hayakawa, Masao*; Tsukada, Takashi; Kaji, Yoshiyuki; Miwa, Yukio*; Ando, Masami*; Nakata, Kiyotomo*
Atsuryoku Gijutsu, 47(4), p.236 - 244, 2009/07
In this study, micro-hardness tests and AFM observations were performed on SUS316L low-carbon austenitic stainless steel pre-strained by cold rolling to investigate its deformation behavior. The following results were obtained. Despite the fact that the same plastic strain was applied, post-tensile test AFM showed narrower slip-band spacing in a reduction in area of 30% cold-rolled specimen than the unrolled specimen. Concentrated slip bands were observed near grain boundaries. Micro-hardness exceeding 300 was found to occur frequently in after tensile test specimens with a reduction in area of 30% or more, particularly at grain boundaries. It is suggested that the nonuniformity of deformation at grain boundaries plays an important role of IGSCC crack propagation mechanism of low-carbon austenitic stainless steel.
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).
Hirose, Takanori; Ando, Masami; Tanigawa, Hiroyasu; Okubo, Nariaki; McDuffee, J. L.*; Heatherly, D. W.*; Sitterson, R. G.*; Stoller, R. E.*; Yamamoto, Takuya*
Fusion Materials Semiannual Progress Report (DOE/ER-0313/46) (Internet), p.72 - 78, 2009/06
This work is being carried out under Annex I of the collaboration on Fusion Materials between the U.S. DOE and the Japan Atomic Energy Agency. The MFE-RB-15J capsule is a part of the Phase-V experiments with the goal of elucidating the effects of helium in fusion structural candidate engineering alloys, and verifying the irradiation response of alloy F82H and its joint for ITER-TBM (Test Blanket Module) application. The target dose of this capsule is 6 dpa at the peak and it will introduce 300 appm of transmutation helium in F82H with additional boron-10. Assembly of the MFE-RB-15J capsule was completed on April, 2008. Irradiation began for MFE-RB-15J with cycle 415, starting June 3, 2008. The MFE-RB-15J experiment was installed in location RB-1A with europium thermal neutron shield. A detailed specimen loading list for the capsule is provided in this report.
Tanigawa, Hiroyasu; Sokolov, M. A.*; Sawahata, Atsushi*; Hashimoto, Naoyuki*; Ando, Masami; Shiba, Kiyoyuki; Enomoto, Masato*; Klueh, R. L.*
Journal of ASTM International (Internet), 6(5), 10 Pages, 2009/05
The master curve (MC) method works when the transition fracture toughness values follow the MC, and once the value is scaled properly, the MC is usually independent of the type of steel or the type of test specimen. This method is very much depending on the assumption that the fracture initiation points are homogeneously distributed and its initiation mechanism is independent on test temperature. The reduced-activation ferritic/martensitic steels (RAFs), such as F82H (Fe-8Cr-2W-0.2V-0.04Ta), has AlO Ta(V,Ti)O composite inclusions, or simple Ta(V)O inclusions, and shows inhomogeneous distribution, and it was revealed that that RAFs which contain Ta could initiate the facture in the different mechanism at lower temperature as the composite inclusions become fragile, and this should be considered when the toughness measured with small size toughness specimen which is usually tested at lower temperature.
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.
Ando, Masami; Tanigawa, Hiroyasu; Wakai, Eiichi; Stoller, R. E.*
Journal of Nuclear Materials, 386-388, p.315 - 318, 2009/04
Radiation hardening and embrittlement due to neutron irradiation around 573 K are the important issues on RAF/M steels. It is expected that the improvement of radiation hardening might be one of effective ways to control the mechanical properties after irradiation. The purposes of this study are to find the condition of heat treatment for minimum of radiation hardening in F82H steel using Neutron/Ion-irradiation and to examine a correlation between tensile property and micro-hardness before/after irradiation. Neutron irradiation was performed in HFIR up to 9 dpa. Ion-irradiation at 573 K was carried out at the TIARA facility of JAEA. For the results of tensile test and hardness test of F82H and F82H heat treatment variants neutron-irradiated at 573 K, all specimens caused radiation hardening. The radiation hardening (H) obtained by hardness test is almost same level, however radiation hardening (YS) of F82H heat treatment variants is smaller than that of F82H.
Ando, Masami; Wakai, Eiichi; Tanigawa, Hiroyasu; Kawasaki, Yasushi
Nihon Kinzoku Gakkai-Shi, 72(10), p.785 - 788, 2008/10
no abstracts in English
Ando, Masami; Wakai, Eiichi; Okubo, Nariaki; Ogiwara, Hiroyuki; Sawai, Tomotsugu; Onuki, Somei*
Nihon Kinzoku Gakkai-Shi, 71(12), p.1107 - 1111, 2007/12
no abstracts in English
Tanigawa, Hiroyasu; Sakasegawa, Hideo; Hashimoto, Naoyuki*; Klueh, R. L.*; Ando, Masami; Sokolov, M. A.*
Journal of Nuclear Materials, 367-370(1), p.42 - 47, 2007/08
It was previously reported that reduced-activation ferritic/martensitic steels (RAFs), such as F82H-IEA and its heat treatment variant, ORNL9Cr-2WVTa, JLF-1 and 2%Ni-doped F82H, showed a variety of changes in ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573K up to 5dpa. These differences could not be interpreted solely as an effect of irradiation hardening caused by dislocation loop formation. To address these observations, the precipitation behavior of the irradiated steels was examined by weight analysis, X-ray diffraction analysis and chemical analysis on extraction residues. The results suggested that irradiation affects precipitation as if it was forced to reach the thermal equilibrium state at irradiation temperature 573K, which usually never be achieved by aging. The details of precipitates in the irradiated RAFs were examined to determine their impact on the mechanical properties, which obtained by tensile, Charpy impact, and bend bar toughness tests. Transmission electron microscopy was performed on thin films and extraction replica specimens to analyze the size distribution, chemical composition and crystal structure of precipitates. It turned out that the hardening level normalized by square root of average packet size showed a linear dependence on the increase of extracted precipitate weight. This dependence suggests that the difference in irradiation hardening between RAFs was caused by the different precipitation behavior on packet, block and prior austenitic grain boundaries during irradiation. The simple Hall-Petch law could be applicable to interpret this dependence. Detailed analytical results will be presented and their interpretation discussed.