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Yamashita, Takuya; Yamashita, Hayato; Nagae, Yuji
Tetsu To Hagane, 105(1), p.96 - 104, 2019/01
Times Cited Count：1 Percentile：6.23(Metallurgy & Metallurgical Engineering)no abstracts in English
Sone, Tomoyuki; Nakagawa, Akinori; Koyama, Hayato; Gunji, Kiyoshi; Nonaka, Kazuharu; Sasaki, Toshiki; Tashiro, Kiyoshi; Yamashita, Toshiyuki
JAEA-Technology 2009-023, 33 Pages, 2009/06
JAEA-Technology-2009-023.pdf:8.11MB
Steam reforming (SR) method consists of the gasification process in which organics are vaporized and decomposed with superheated steam and the oxidation process in which vaporized organics are decomposed by oxidizing reaction with heated air. 2,500L of waste TBP/n-dodecane contaminated with uranium was treated using the demonstration scale steam reforming system to examine the performance of the system. Results obtained in this study show that the temperature in the SR system was controlled under the self-regulation temperature, the concentration of CO and NOx in the off-gas were controlled less than 100ppm and 250ppm respectively, the distribution ratio of uranium to off-gas treatment system was under 0.12% and the gasification ratio of waste TBP was more than 99%. This long-term waste treatment test has demonstrated that the SR system can safely and effectively reduce the volume of the waste.
Koyama, Hayato; Sone, Tomoyuki; Sasaki, Toshiki; Yamashita, Toshiyuki
no journal, ,
no abstracts in English
Yamashita, Hayato; Yamashita, Takuya; Onizawa, Takashi; Nagae, Yuji; Yamamoto, Kenji*; Shudo, Shingo*; Kawasaki, Kenji*; Kubo, Koji*
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no abstracts in English
Okuda, Takahiro; Yamashita, Hayato; Toyota, Kodai; Shimomura, Kenta; Onizawa, Takashi; Kato, Shoichi
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This study describes the setting of the material property equations of 316FR steel at an extremely high temperature which can be applied to severe accident conditions of generation IV fast reactors. 316FR steel will be applied to structural materials, e.g. reactor vessel, in the generation IV fast reactors. After the severe accident in Fukushima Daiichi Nuclear Power Plants, the evaluation of structural integrity was found to be very important severe accident condition. The development of the generation IV fast reactors requires the material properties of 316FR steel at the extremely high temperature. However, such data has not been acquired. Therefore, tensile and creep tests were carried out in the temperature range over 700
C for 316FR steel. Based on the acquired data from the tests, the equations that can evaluate the material properties of 316FR steel at the extremely high temperature were set up. They are an elasto-plastic stress-strain equation, a creep rupture equation and a creep strain equation.
Yamashita, Hayato
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no abstracts in English
Yamashita, Hayato; Onizawa, Takashi; Takaya, Shigeru; Enuma, Yasuhiro
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Yamashita, Hayato; Onizawa, Takashi; Takaya, Shigeru; Enuma, Yasuhiro; Wakai, Takashi; Kato, Shoichi; Suzuki, Akihiro*
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This study proposes an appropriate method to convert the load parameter in small punch creep (SPC) tests to that in uniaxial creep tests for Mod.9Cr-1Mo steel at operating temperature of future fast reactors. It is impossible to simply compare the results of SPC tests to those of uniaxial creep tests, because the load parameters in those creep tests are load and stress, respectively. Conventionally, load/stress conversion coefficient (F/
) has been employed to compare the results of SPC and uniaxial creep tests. Though the applicability of F/ has been demonstrated in the temperature region considerably higher than 550C that is the operating temperature of future fast reactors, the applicability has not been demonstrated at 550C. Therefore, this paper studies an appropriate method to obtain F/. F/ calculated from finite element analysis (FEA) and that based on fracture ductility was too small to convert SPC results. In contrast, it was clarified that conversion was possible using F/ based on Arrhenius-type equation.
Yamamoto, Kenji*; Yamashita, Hayato; Onizawa, Takashi
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Yamamoto, Kenji*; Yamashita, Hayato; Onizawa, Takashi; Kimura, Kazuhiro*; Sawada, Kota*; Hongo, Hiromichi*
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316FR steel is a candidate material for the main components of the fast reactor, and its design allowable stress is specified in Japan Society of Mechanical Engineers fast reactor standard. The reactor will be larger in size. For the design, welding is also required at a part to be a high-temperature and high-stress. Therefore, allowable stress for the weld joint is required. In the evaluation of welded joints, it is necessary to consider the characteristics of the weld metal and heat affected zone in addition to the base material. In this paper, by comparing the strength properties of 316FR steel base metal and weld joint, it was examined whether it is necessary to specify the allowable stress of the weld joint in the restriction of the primary stress of the JSME fast reactor standard.
Oshikiri, Masato*; Kubo, Koji*; Ando, Masanori; Yamashita, Hayato
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no abstracts in English
Oshikiri, Masato*; Kubo, Koji*; Ando, Masanori; Yamashita, Hayato
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no abstracts in English
