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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
Times Cited Count：4 Percentile：32.74(Nuclear Science & Technology)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.
Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
Journal of Plasma and Fusion Research SERIES, Vol.11, p.69 - 72, 2015/03
Water-cooled blanket is an attractive concept for its compactness and its compatibility with the conventional technologies for PWR. For blanket application, the structural material is required to be as thin as possible for tritium breeding. On the other hand, it is also required the pressure tightness to withstand 15 MPa of internal pressure. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. The effects of water flow and DO in the test water on corrosion properties were investigated using rotating disk specimen in autoclave. In summary, the weight loss by flowing was occurred except for test with DO 8 ppm, and it was more pronounced at lower DO concentration. Since Fe
O
was observed on the specimen of small weight change, and the iron-poor layer thickness increased with decreasing the specimen weight, it seemed that the formation of FeO was effective for the suppression of weight loss.
Kanai, Akihiko*; Kasada, Ryuta*; Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio; Konishi, Satoshi*
Journal of Nuclear Materials, 455(1-3), p.431 - 435, 2014/12
Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)Kanai, Akihiko*; Park, C.*; Noborio, Kazuyuki*; Kasada, Ryuta*; Konishi, Satoshi*; Hirose, Takanori; Nozawa, Takashi; Tanigawa, Hiroyasu
Fusion Engineering and Design, 89(7-8), p.1653 - 1657, 2014/10
Times Cited Count：5 Percentile：36.65(Nuclear Science & Technology)Sato, Satoshi; Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio; Ochiai, Kentaro; Konno, Chikara
Fusion Engineering and Design, 89(9-10), p.1984 - 1988, 2014/10
Times Cited Count：2 Percentile：16.31(Nuclear Science & Technology)In order to evaluate nuclear properties of the ITER JA WCCB-TBM (Water Cooled Ceramic Breeder Test Blanket Module) and ensure that the design conforms to the nuclear regulation for licensing, nuclear analyses have been performed for the WCCB-TBM including flame, shield, pipe-forest, bio-shield and AEU (Ancillary Equipment Unit). Nuclear analyses are performed with the Monte Carlo code MCNP5.14, activation code ACT-4 and Fusion Evaluated Nuclear Data Library FENDL-2.1. MCNP geometry input data of the TBM is created from CAD data with the automatic conversion code GEOMIT, and other geometry input data is created by manually. By adopting the dog-leg gaps, decay 
-ray dose rate can be drastically reduced and hands-on access is possible for shield. Detailed calculation results will be presented in this symposium.
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.
Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10
Times Cited Count：21 Percentile：83.69(Nuclear Science & Technology)The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.
Hirose, Takanori; Nozawa, Takashi; Stoller, R. E.*; Hamaguchi, Dai; Sakasegawa, Hideo; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Enoeda, Mikio; Kato, Yutai*; Snead, L. L.*
Fusion Engineering and Design, 89(7-8), p.1595 - 1599, 2014/10
Times Cited Count：47 Percentile：96.55(Nuclear Science & Technology)The material properties, focusing on the properties used for design analysis were re-assessed and newly investigated for various heats including F82H-IEA. Moreover, irradiation effects on those properties were studied in this work. As for thermal properties, thermal conductivity that has significant impacts on the thermo-hydraulic properties of the blanket was investigated on several heats of F82H including F82H-IEA. According to the measurements, the thermal conductivity falls in the range 28.3
1.1 W/m/K at 293 K. Although this is comparable with that of the other ferritic/martensitic steels, it is 20% lower than the published value for F82H-IEA. The re-assessment on the published value revealed that the thermal diffusivity was over-estimated. As for irradiation effects on the physical properties, electric resistivity was measured after irradiation up to 6 dpa at 573 K and 673 K. The reduction of resistivity in F82H and its welds were 3% and 6%, respectively.
Ando, Masanori; Hirose, Yuichi*; Karato, Takanori*; Watanabe, Sota*; Inoue, Osamu*; Kawasaki, Nobuchika; Enuma, Yasuhiro*
Journal of Pressure Vessel Technology, 136(4), p.041406_1 - 041406_10, 2014/08
Times Cited Count：9 Percentile：43.32(Engineering, Mechanical)To compare and assess the creep-fatigue life evaluation methods for stress concentration point, a series of creep-fatigue test was performed with notched specimens made of Mod.9Cr-1Mo steel. Mechanical creep-fatigue tests and thermal creep-fatigue test were performed. A series of Finite Element Analysis was also carried out to predict the number of cycles to failure by the several creep-fatigue life evaluation methods. Then these predictions were compared with the test results. Several types of evaluation methods such are stress redistribution locus (SRL) method, simple elastic follow-up method and the methods described in the JSME FRs code were applied. Through the comparisons, it was appeared that SRL method gave rational conservative prediction of the creep-fatigue life for all conditions tested in this study. The JSME FRs code gave an evaluation over 70 times conservative lives comparing with the test results.
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 500
C 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; Someya, Yoji; Tanigawa, Hisashi; Suzuki, Satoshi
Yosetsu Gakkai-Shi, 83(1), p.70 - 77, 2014/01
no abstracts in English
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
Times Cited Count：9 Percentile：57.11(Materials Science, Multidisciplinary)Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; et al.
Fusion Engineering and Design, 87(7-8), p.1363 - 1369, 2012/08
Times Cited Count：36 Percentile：92.34(Nuclear Science & Technology)The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. Fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.
Seki, Yohji; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ezato, Koichiro; Enoeda, Mikio; Sakamoto, Kensaku
Dai-17-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.265 - 266, 2012/06
In the case of a water cooled ceramic breeder in a blanket, pebbles of a ceramic tritium breeder are packed in a container constituted by a partition plate. Helium purge gas is applied as a transport fluid in a tritium recovery system. It is of importance to build database of a pressure drop as part of a design of the tritium recovery system. In this experimental study, the pressure drops of He gas through pebble bed were measured within the wide range of a flow rate up to 100 L/min. The results indicate that a laminar flow is dominant and the pressure drop was correctly predicted by the empirical equation within a part of flow rate. Reliability of prediction ability of pressure drop was established by this experiment within the flow rate which is less than 60 L/min. Moreover, this paper describes that slight difference between the experimental result and the empirical equation within a range of flow rate from 60 L/min to 100 L/min.
Shiba, Kiyoyuki; Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro
Fusion Engineering and Design, 86(12), p.2895 - 2899, 2011/12
Times Cited Count：43 Percentile：94.34(Nuclear Science & Technology)Aging properties of reduced activation ferritic/martensitic steel F82H was researched at temperature ranging from 400C to 650C up to 100,000 hr. Microstructure, tensile, and Charpy properties were carried out. Laves was found at temperatures between 550 and 650C and M
C carbides were found at the temperatures between 500 and 600C over 10,000 hr. These precipitates caused degradation in toughness, especially at temperatures ranging from 550C to 650C. Tensile properties do not have serious aging effect, except for 650C, which caused large softening even after 10.000 hr. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in DBTT caused by the large Laves phase precipitation at grain boundary was observed in the 650C aging. Laves precipitates at grain boundary also degrades the USE of the aged materials. These aging test results provide F82H can be used up to 30,000 hr at 550C.
Seki, Yohji; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
Proceedings of Plasma Conference 2011 (PLASMA 2011) (CD-ROM), 2 Pages, 2011/11
Development of test blanket module (TBM) with a water cooled solid breeder is being performed as the primary candidate of ITER-TBM of Japan. Prior to the installation of each TBM, it is necessary to develop the capability of the prediction analyses of all essential functions of the blanket to validate the analyses tools by the TBM. Especially the prediction tool of tritium concentration in the blanket system is one of the most important issues to control tritium recovery. From this view point, this paper discusses the flow phenomena and the tritium transport of the helium purge gas in the pebble bed. By prediction of purge gas flow using a numerical simulation, the result indicates tritium concentration depended on the position of the breeder layer. Namely, the large concentration still remains near the wall with approaching to an outlet.
Hirose, Takanori; Tanigawa, Hisashi; Yoshikawa, Akira; Seki, Yohji; Tsuru, Daigo; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato
Fusion Engineering and Design, 86(9-11), p.2265 - 2268, 2011/10
Times Cited Count：5 Percentile：38.41(Nuclear Science & Technology)As one of the most important fabrication technologies of the WCCB TBM, Hot Isostatic Pressing (HIP) joining technology was selected to fabricate the first wall with built-in cooling channel structure made of reduced activation martensitic/ferritic steel, F82H. By using developed HIP technology, a real scale TBM first wall mockup was successfully fabricated. High heat flux test of the fabricated mockup showed the feasibility to with the equivalent conditions of the WCCB TBM operation. The breeder pebble box was successfully fabricated with thin wall cooling pipes and thin plate sleds by Laser welding. With respect to the side walls with built in cooling channels were also fabricated using drilling technology. Assembling of the first wall and side walls is one of the critical fabrication processes of the fabrication of the TBM structure. By using a F82H first wall mockup and side wall mockups, assembling process was demonstrated successfully by Electron Beam welding.
Okubo, Nariaki; Sokolov, M. A.*; Tanigawa, Hiroyasu; Hirose, Takanori; Jitsukawa, Shiro; Sawai, Tomotsugu; Odette, G. R.*; Stoller, R. E.*
Journal of Nuclear Materials, 417(1-3), p.112 - 114, 2011/10
Times Cited Count：10 Percentile：60.56(Materials Science, Multidisciplinary)Irradiation hardening and fracture toughness of reduced-activation ferritic/martensitic steel F82H after irradiation were investigated with a focus on changing the fracture toughness transition temperature as a result of several heat treatments. The specimens were standard F82H-IEA (IEA), F82H-IEA with several heat treatments (Mod1 series) and a higher tantalum containing (0.1%) heat of F82H (Mod3). The specimens were irradiated up to 18 dpa at 300 C in High Flux Isotope Reactor under a collaborative research program between JAEA/US-DOE. The results of hardness tests showed that irradiation hardening of IEA was comparable with that of Mod3. However, the fracture toughness transition temperature of Mod3 was lower than that of IEA. The transition temperature of Mod1 was also lower than that of the IEA heat. These results suggest that tightening of specifications on the heat treatment condition and modification of the minor alloying elements seem to be effective to reduce the fracture toughness transition temperature after irradiation.
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
Times Cited Count：18 Percentile：78.57(Materials Science, Multidisciplinary)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.
Seki, Yohji; Onishi, Yoichi*; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ozu, Akira; Ezato, Koichiro; Tsuru, Daigo; Suzuki, Satoshi; Yokoyama, Kenji; et al.
Progress in Nuclear Science and Technology (Internet), 2, p.139 - 142, 2011/10
R&D of a test blanket module (TBM) with a water-cooled solid breeder has been performed for ITER. For our design, the temperature of a coolant pressurized up to 15 MPa is designed as 598 K in an outlet of the TBM, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. The Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the pressure drop and flow rates in the channels of the side wall. It results the inhomogeneous flow rates in each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mock-up.
-US
Sugiyama, Kiyohiro*; Hirose, Yusuke*; Enoki, Kentaro*; Ikeda, Shugo*; Yamamoto, Etsuji; Tateiwa, Naoyuki; Haga, Yoshinori; Kida, Takanori*; Hagiwara, Masayuki*; Kindo, Koichi*; et al.
Journal of the Physical Society of Japan, 80(Suppl.A), p.SA104_1 - SA104_3, 2011/07
Times Cited Count：3 Percentile：27.36(Physics, Multidisciplinary)