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Ukai, Shigeharu*; Yano, Yasuhide; Inoue, Toshihiko; Sowa, Takashi*
Materials Science & Engineering A, 812, p.141076_1 - 141076_11, 2021/04
Times Cited Count：12 Percentile：73.14(Nanoscience & Nanotechnology)FeCrAl oxide dispersion strengthened alloys are promising materials for accident tolerant fuels for light water reactors (LWRs). In these alloys, Al and Cr are key elements with important synergistic effects: enhancement of the formation of oxidation-resistant Al
O
phase by Cr addition and suppression of the formation of the embrittling Cr-rich 
' phase by Al addition. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The Al and Cr contents were systematically varied from 9-16 at.% and 10-17 at.%, respectively, and tensile tests were conducted at 298 K, 573 K and 973 K in the as-annealed condition. The solid solution strengthening increased linearly, 20 MPa per 1 at.% Al and 5 MPa per 1 at.% Cr, at the typical LWR operational temperature of 573 K. The conventional Fleischer-Friedel and Labusch theories cannot explain this level of solid-solution strengthening. It was shown that Suzuki's double kink theory for screw dislocations reasonably predicts the solid solution strengthening by Al and Cr as well as the inverse dependency on the absolute temperature and linear dependency on the Al and Cr content.
Ito, Chikara; Maeda, Shigetaka; Inoue, Toshihiko; Tomita, Hideki*; Iguchi, Tetsuo*
Radiation Protection, 40(6), p.491 - 495, 2020/11
A highly accurate and precise technique for measurement of the 
Nb(n,n')
Nb reaction rate was established for the material surveillance tests, etc. in fast reactors. The self-absorption effect on the measurement of the characteristic X-rays emitted by Nb was decreased by the solution and evaporation to dryness of niobium dosimeter. A highly precise count of the number of Nb atoms was obtained by measuring the niobium solution concentration using inductively coupled plasma mass spectrometry. X-rays of Nb were measured accurately by means of comparing the X-ray intensity of irradiated niobium solution with that of the solution in which stable Nb was added. The difference between both intensities indicates the effect of 
Ta, which is generated from an impurity tantalum, and the intensity of X-rays from Nb was evaluated. Measurement error of the Nb(n,n')Nb reaction rate was reduced to be less than 4%, which was equivalent to the other reaction rate errors of dosimeters used for Joyo dosimetry. In addition, an advanced technique using Resonance Ionization Mass Spectrometry was proposed for the precise measurement of Nb yield, and Nb will be resonance-ionized selectively by discriminating the hyperfine splitting of the atomic energy levels between Nb and Nb at high resolution.
Yano, Yasuhide; Sekio, Yoshihiro; Tanno, Takashi; Kato, Shoichi; Inoue, Toshihiko; Oka, Hiroshi; Otsuka, Satoshi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; et al.
Journal of Nuclear Materials, 516, p.347 - 353, 2019/04
Times Cited Count：15 Percentile：86.05(Materials Science, Multidisciplinary)9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 
C, which surpassed that of heat-resistant austenitic steels although creep rupture strength of tempered martensitic steels is generally lower than that of austenitic steels at high temperatures. The measured creep rupture strength of 9Cr-ODS steel claddings at 1000 C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 C. This superior strength seemed to be owing to transformation of the matrix from the -phase to the 
-phase. The transient burst strengths for 9Cr-ODS steel were much higher than those for 11Cr-ferritic/martensitic steel (PNC-FMS). Cumulative damage fraction analyses suggested that the life fraction rule can be used for the rupture life prediction of 9Cr-ODS steel and PNC-FMS claddings in the transient and accidental events with a certain accuracy.
Inoue, Toshihiko; Sekio, Yoshihiro; Watanabe, Hideo*
Materia, 58(2), P. 92, 2019/02
For the evaluation of irradiated segregation behavior, Austenite-based stainless steel for the fast reactor, during irradiation was evaluated by utilizing TIARA facility (Irradiate temperature: 600 C, Dose: 100 dpa) was observed by analytical electron microscope (JEM-ARM20FC). As a result of observation, the large-size void is observed in irradiation area, and MX segregation (containing Niobium) is not observed. In un-irradiation area the MX segregation is observed. And it is observed conspicuously that Nickel is segregation on the void surface. By the latest high-performance TEM utilization, these phenomenon are able to visualize. It is expected for the clarification of the irradiation damage and mechanism of void swelling, by the analyzing these phenomenon utilization with the latest high-performance TEM utilization.
Ojima, Mayumi*; Shiro, Ayumi*; Suzuki, Hiroshi; Inoue, Junya*; Shobu, Takahisa; Xu, P. G.; Akita, Koichi; Nambu, Shoichi*; Koseki, Toshihiko*
Zairyo, 66(6), p.420 - 426, 2017/06
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.
Journal of Nuclear Materials, 487, p.229 - 237, 2017/04
Times Cited Count：37 Percentile：96.77(Materials Science, Multidisciplinary)Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200C. This degradation was attributed to grain boundary sliding deformation with /
transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 C unlike the other tested materials. Present study includes the result of "R&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
Sugaya, Yuki; Sakazume, Yoshinori; Akutsu, Hideyuki; Inoue, Toshihiko; Yoshimochi, Hiroshi; Sato, Soichi; Koyama, Tomozo; Nakayama, Shinichi
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 8 Pages, 2017/00
The Japan Atomic Energy Agency has been developing the research and development facilities, "Okuma Analysis and Research Center", in order to ascertain the properties of radioactive wastes and fuel debris towards the decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station. This paper outlines the concept of "Laboratory-1" which will analyze low and medium level samples in the Okuma Analysis and Research Center with a focus on the research plan.
Ito, Masayasu; Ogawa, Miho; Inoue, Toshihiko; Yoshimochi, Hiroshi; Koyama, Shinichi; Koyama, Tomozo; Nakayama, Shinichi
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 7 Pages, 2017/00
Laboratory-2 of the Okuma Analysis and Research Center will be used for the technological development of techniques to treat and dispose fuel debris, etc. The specific analytical content and its importance has been discussed by an experts committee in FY 2016. The committee regarded fuel debris retrieval and criticality control related topics as the most important content. As a result, it will be a priority to introduce equipment to perform examination such as shape and size measurement, compositional and nuclide analysis, hardness and toughness test, and radiation dose rate measurement. In addition, since sample will have high dose rates (1 Sv/h or more) at the time of reception, hot cells with enough radiation shielding ability will be used. In the hot cell, the pre-processing will be performed, such as cutting and dissolution of samples. Processed samples will be examined in concrete cells, steel cells, glove boxes and fume hoods. Detail design of Laboratory-2 started on FY 2017.
Inoue, Toshihiko; Ogawa, Miho; Sakazume, Yoshinori; Yoshimochi, Hiroshi; Sato, Soichi; Koyama, Shinichi; Koyama, Tomozo; Nakayama, Shinichi
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 7 Pages, 2017/00
Decommissioning of TEPCO's 1F is in progress according to the Roadmap. The Roadmap assigned the construction of a hot laboratory and analysis to the JAEA. The hot laboratory, Okuma Analysis and Research Center consists of the three buildings; Administrative building, the Laboratory-1 and Laboratory-2. The Laboratory-1 and Laboratory-2 are hot laboratories. Laboratory-1 is for radiometric analysis of low and medium level radioactive rubble and secondary wastes. The license of the Laboratory-1's implementation was approved by The Secretariat of the Nuclear Regulation Authority and the construction started in April 2017 and plans an operational start in 2020. Laboratory-2 provides concrete cells, steel cells for the analysis of the fuel debris and high level radioactive rubble. The Laboratory-2's major analysis items is reviewed by review meeting organized of cognoscente.
Osaka, Masahiko; Donomae, Takako; Ichikawa, Shoichi; Sasaki, Shinji; Ishimi, Akihiro; Inoue, Toshihiko; Sekio, Yoshihiro; Miwa, Shuhei; Onishi, Takashi; Asaka, Takeo; et al.
Proceedings of 1st Asian Nuclear Fuel Conference (ANFC), 2 Pages, 2012/03
Support system for training and education of future expert in hot laboratories of Oarai-JAEA, named FEETS, is presented. The system has been established based on research results on both characterization of Oarai hot laboratory and user-needs. Various programs under FEETS are also introduced.
Ojima, Mayumi*; Inoue, Junya*; Nambu, Shoichi*; Xu, P. G.; Akita, Koichi; Suzuki, Hiroshi; Koseki, Toshihiko*
Scripta Materialia, 66(3-4), p.139 - 142, 2012/02
Times Cited Count：78 Percentile：94.99(Nanoscience & Nanotechnology)Inoue, Toshihiko; Yamagata, Ichiro; Asaka, Takeo
Nihon Genshiryoku Gakkai-Shi ATOMO
, 53(9), p.638 - 642, 2011/09
This paper presents the properties and the development of modified 316 steel. The core material is required to high-temperature strength caused by high power density and FP gas, swelling resistance caused by irradiation damage, and corrosion resistance caused by coolant sodium and FCCI. In the improvement of the modified 316 stainless steel, screening test conducted to improve high-temperature strength and swelling resistance. Optimizing a slight amount of addition element and cold working, the modified 316 steel was improved the high-temperature strength and swelling resistance in both. And under irradiated and FCCI conditions, these properties were tested. The modified 316 stainless steel uses 44,000 pins as fast reactor fuel pin in JOYO. These results show that the steel exhibits excellent characteristics in creep rupture strength and swelling resistance.
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
JAEA-Research-2010-019-01.pdf:48.47MBJAEA-Research-2010-019-02.pdf:19.4MB
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.
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
Times Cited Count：137 Percentile：97.72(Physics, Fluids & Plasmas)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).
Kaito, Takeji; Otsuka, Satoshi; Inoue, Masaki; Asayama, Tai; Uwaba, Tomoyuki; Mizuta, Shunji; Ukai, Shigeharu*; Furukawa, Tomohiro; Ito, Chikara; Kagota, Eiichi; et al.
Journal of Nuclear Materials, 386-388, p.294 - 298, 2009/04
Times Cited Count：32 Percentile：88.55(Materials Science, Multidisciplinary)In order to examine irradiation effect on creep rupture strength of Oxide Dispersion Strengthened (ODS) steel claddings, the in-pile creep rupture test was conducted using Material Testing Rig with Temperature Control (MARICO)-2 in the experimental fast reactor JOYO. Fourteen creep rupture events were successfully detected by the temperature change in each capsule and the -ray spectrometry of the cover gas. Time to creep ruptures of six ODS steel specimens were identified by means of Laser Resonance Ionization Mass Spectrometry (RIMS), and no irradiation effect on creep rupture strength was confirmed within the irradiation condition in the MARICO-2 test.
Ushigusa, Kenkichi; Seki, Masahiro; Ninomiya, Hiromasa; Norimatsu, Takayoshi*; Kamada, Yutaka; Mori, Masahiro; Okuno, Kiyoshi; Shibanuma, Kiyoshi; Inoue, Takashi; Sakamoto, Keishi; et al.
Genshiryoku Handobukku, p.906 - 1029, 2007/11
no abstracts in English
Mori, Masahiro; Shoji, Teruaki; Araki, Masanori; Saito, Keiji*; Senda, Ikuo; Omori, Junji*; Sato, Shinichi*; Inoue, Takashi; Ono, Isamu*; Kataoka, Takahiro*; et al.
Nihon Genshiryoku Gakkai-Shi, 44(1), p.16 - 89, 2002/01
no abstracts in English
Hayashi, Takumi; Nakamura, Hirofumi; Konishi, Satoshi; Yamanishi, Toshihiko; Inoue, Masahiko*; Hirata, Kazuhiro*; Okuno, Kenji; Naruse, Yuji; Anderson, J. L.*; Barnes, J. W.*; et al.
JAERI-M 93-081, 35 Pages, 1993/03
no abstracts in English
Inoue, Masahiko; Konishi, Satoshi; Yamanishi, Toshihiko; Ohira, Shigeru; Watanabe, Tetsuo*; Okuno, Kenji; Naruse, Yuji; Sherman, R. H.*; Barnes, J. W.*; Bartlit, J. R.*; et al.
Fusion Technology, 21(2P2), p.293 - 298, 1992/03
Times Cited Count：3 Percentile：35.32(Nuclear Science & Technology)no abstracts in English
Inoue, Toshihiko; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji
no journal, ,
no abstracts in English
