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Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi
no journal, ,
JAEA has started to develop a technology which can be applicable to high radioactive special nuclear materials such as next-generation fuel cycle products. We have been developed Non-destructive assay system Active-N as a test equipment which utilizes D-T neutron generator. In a system for Differential Die-Away (DDA) method which is tested in Active-N, it is important to evaluate neutron flux to check the performance of the system. In this research, we have evaluated neutron flux in a system for Fast Neutron Direct Interrogation method which is a kind of DDA method by activation method and Monte Carlo simulation by using PHITS.
Terunuma, Hirotaka; Tanaka, Kiwamu; Kabumoto, Hiroshi; Haginoya, Masashi; Sano, Naruto; Takahashi, Masatomi; Hoshino, Masato; Aoki, Isao; Asazuma, Shinichiro
no journal, ,
no abstracts in English
generation by reaction with Ru-La-Na mix nitrates and raw materials for vitrificationNagai, Takayuki; Kobayashi, Hidekazu; Okamoto, Yoshihiro; Sato, Nobuaki*; Inose, Takehiko*; Sato, Seiichi*; Hatakeyama, Kiyoshi*; Seki, Katsumi*
no journal, ,
It is thought that a generated ruthenium compound grows from a high level radioactive liquid waste into RuO crystal by reacting to raw materials for the vitrification process. In this study, the generation reaction to RuO was confirmed by heating Ru-La-Na mix nitrates and the raw materials.
Tsuchiya, Harufumi; Kitatani, Fumito; Kureta, Masatoshi; Maeda, Makoto
no journal, ,
no abstracts in English
Ozu, Akira; Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Kureta, Masatoshi
no journal, ,
no abstracts in English
Tada, Kenichi; Nagaya, Yasunobu
no journal, ,
JAEA has been developing the nuclear data processing code FRENDY (FRom Evaluated Nuclear Data librarY to any application). In this presentation, construction of the probability table in the unresolved resonance region is described.
Kikuchi, Takeo; Tada, Kenichi; Suyama, Kenya
no journal, ,
To estimate the prediction accuracy of the integrated burn up analysis code system SWAT4, we compared the calculation results of SWAT4 and the PIE data of the BWR fuel which was measured by JAERI in 1990s. Comparison results are indicated that the C/E value of major heavy nuclei, e.g., U and Pu, is approximately 1.0. The calculation results are also indicated that some fission products, e.g., Sm, have the larger difference.
Zaima, Naoki; Naganuma, Masaki; Sakao, Ryota
no journal, ,
no abstracts in English
Iwai, Yasunori; Kondo, Akiko*; Edao, Yuki; Sato, Katsumi; Kubo, Hitoshi*; Oshima, Yusuke*
no journal, ,
Effect of halogenated gas on detritiation efficiency of the detritiation system has been investigated taking an event of off normal event such as fire into consideration. Concerning the activity of platinum catalyst for oxidation of tritium, we have evaluated the steep decrease in activity of platinum catalyst in the presence of halogenated gas. In order to avoid the steep decrease in activity, a noble catalyst alloyed with platinum and palladium showed an outstanding proof. In addition, the halogenated acid produced over catalyst surface affects the activity of catalyst. As for water absorber, a molecular sieve decreased its water absorbing capacity in the presence of halogenated gas.
Furutachi, Naoya; Minato, Futoshi; Iwamoto, Osamu
no journal, ,
To establish the nuclear transmutation system for the long-lived fission products (LLFPs), it is desired to improve precision of the simulation calculation for the transmutation system. To achieve this, nuclear data of various nuclei produced via the nuclear transmutation of LLFPs are also important. However, it is expected that unstable nuclei with no available experimental data are produced via the nuclear transmutation. One of the physical quantity that is very difficult to predict with no experimental data is the thermal neutron capture cross section. The thermal neutron capture cross section is dominated by the energy and width of the first resonance, and slight variation of them can change the thermal neutron capture cross section drastically. While it is very difficult to determine them with high precision, it is known that a resonance width follows Porter-Thomas distribution because of complexity and randomness of a nuclear structure, and a resonance spacing follows Wigner distribution. In this work, we calculate the thermal neutron capture cross section by using the statistical property of the resonance parameters with Monte Carlo method. The calculation result is obtained as a probability distribution of the thermal neutron capture cross section. We calculated approximately 250 nuclei that have experimental data, and found that the dispersion of the experimental data is well explained by the calculated probability distribution.
Matsuda, Norihiro; Sato, Tatsuhiko; Niita, Koji*; Suyama, Kenya
no journal, ,
no abstracts in English
Tanigawa, Hisashi; Gwon, H.; Kawamura, Yoshinori
no journal, ,
JAEA is developing a water-cooled ceramic breeder blanket. For the blanket strength and pressure integrity are assessed. The largest stress appears in the first wall region due to the surface heat and neutron loads. Under conditions with the thermal loads and cooling water pressure, stress in the first wall is analyzed with reference to ASME Boiler and Pressure Vessel Code. Limitations related to the primary/secondary stresses and strain are considered. Material data necessary for the assessment is summarized, and then status of preparation is studied for a structural material of reduced activation ferritic/martensitic steel, F82H. The summarized date of F82H is compared with standardized 9Cr-1Mo-V.
Amano, Katsunori; Enuma, Yasuhiro; Futagami, Satoshi; Ushiki, Hiroshi*; Kawamura, Masaya*; Ichihara, Takashi*
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no abstracts in English
Ishikawa, Makoto; Yokoyama, Kenji; Numata, Kazuyuki; Maruyama, Shuhei; Takeda, Toshikazu*
no journal, ,
By utilizing the MA-related measurement data, a study to adjust the JENDL-4.0-based cross sections was performed. As a result, it was obtained that the uncertainty of MA-related reactor core parameters induced from the nuclear data uncertainty could be greatly reduced.
Ikeda, Yoshitaka; JT-60SA Team
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no abstracts in English
Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Morita, Koji*
no journal, ,
After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Specifically, emergency power generators are not necessary and the decay heat can be passively removed for a long time, even forever if the heat removal capacity of the RCCS is sufficient. Moreover, we can also define the experimental conditions on radiation and natural convection for the scale-down heat removal test facility.
Mimura, Ryuji; Yokozuka, Yuta; Nemoto, Koichi; Shiraishi, Kunio
no journal, ,
no abstracts in English
Nozawa, Takashi; Yu, J.-H.*; Park, J.-S.*; Tanigawa, Hiroyasu
no journal, ,
Silicon carbide composites are one of promising materials due to pronounced low radioactivity and excellent radiation resistivity. So far, it was proven that the edge effect of the specimen decreased composite strength depending on the specimen size for the case of off-axial tensile tests. This study then aims to evaluate fracture behavior of SiC/SiC tube materials without any edges. For that purpose, several fabric architecture composites were evaluated by expanding plug burst test to identify the fundamental data of anisotropy of the hoop strength of the composite.
Hirose, Takanori; Sakasegawa, Hideo; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Kawamura, Yoshinori
no journal, ,
Material database for reduced activation ferritic/martensitic steel, F82H have been accumulated for design study on tritium breeding blanket. This study reports fatigue properties of F82H including effects of test temperature and test atmosphere to establish design fatigue curve. It is found that temperature effects on fatigue lifetime is significant above 450
C. As for effects of test atmosphere, lifetime in the vacuum was 5 times higher than that in the air at 550C.
Ishimi, Akihiro; Katsuyama, Kozo; Akasaka, Naoaki; Misawa, Susumu*
no journal, ,
no abstracts in English