Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Sasaki, Takayuki*; Akimoto, Yuji*; Seki, Kotaro; Nagano, Misato*; Ishimori, Kenichiro; Ueno, Takashi; Kameo, Yutaka
no journal, ,
no abstracts in English
-concrete system under randomizationUeki, Taro
no journal, ,
Analysis framework under indeterminate material distribution is investigated for the Monte Carlo (MC) criticality calculation of continuously mixed media formed via molten core concrete interaction. Randomized Weierstrass functions (RWF) are utilized to represent the volume fractions of constituent materials. The possibility of several percent fluctuation of effective multiplication factor is shown by the MC simulation with delta-tracking.
Sato, Yuki; Kishimoto, Aya*; Kaburagi, Masaaki; Kataoka, Jun*; Torii, Tatsuo
no journal, ,
no abstracts in English
Edao, Yuki; Iwai, Yasunori; Sato, Katsumi; Hayashi, Takumi
no journal, ,
no abstracts in English
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.
Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Tsujimoto, Kazufumi
no journal, ,
Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuel which has large decay heat. So it is necessary to consider the increase of the core temperature when the core cooling system is stopped. This change of the core temperature was evaluated by thermal conduction analysis. It was found that the impact of thermal insulation in the empty rectangular lattice matrix area was large. Testing equipment was fabricated to verify the accuracy of calculation model for the empty lattice matrix which was the free convection model of sealed fluid. By using this equipment, thermal distribution and one dimensional heat flow through the lattice were measured. It was observed that the actual equivalent thermal conductivity in the lattice was larger than the free convection model. It was also confirmed that the insertion of the aluminum block into the empty lattice could achieve the higher equivalent thermal conductivity. These results could be the useful data for the thermal conduction analysis.
Arai, Masaji; Takino, Kazuo
no journal, ,
The Department of Research Reactor and Tandem Accelerator began to discuss basic design concept of the new multipurpose research reactor succeeding to JRR-3.
Yamano, Hidemasa; Kawasaki, Nobuchika; Kubo, Shigenobu
no journal, ,
Tentative external hazard conditions assumed for a next-generation SFR design in 2011 were not reasonable because of an envelope range covering severe conditions. This report proposed a reasonable definition method of external hazard conditions and defined specific design basis and design extension conditions against various external hazards.
Gwon, H.; Tanigawa, Hisashi; Nakajima, Motoki; Hirose, Takanori; Kawamura, Yoshinori
no journal, ,
It is expected that the neutron wall loading in DEMO is larger than that in ITER, over 0.78 MW/m
. It is concerned that the decay heat due to the large neutron wall loading will lead to excessive temperature rising in blanket. In present study, the thermal response characteristics of blanket caused by the decay heat under LOCA were evaluated. In addition we considered how to effectively mitigate the excessive temperature rising based on the evaluation results.
Okano, Fuminori; JT-60SA Team
no journal, ,
no abstracts in English
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*
no journal, ,
no abstracts in English
Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Tsujimoto, Kazufumi
no journal, ,
This presentation shows fabrication and test results of a testing facility for fuel cooling that is a component of the testing facility for remote-handling of highly-radioactive MA fuels in the transmutation physics experimental facility (TEF-P) planned in the J-PARC. Evaluation formula of pressure drop and temperature increase used in the design of TEF-P was validated by the test, and, feasibility of cooling concept was confirmed.