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Yokoyama, Kenji; Yamamoto, Akio*
no journal, ,
no abstracts in English
Tsukimori, Kazuyuki; Ando, Masanori; Yada, Hiroki; Anoda, Yoshinari*; Ichimiya, Masakazu*; Uno, Masayoshi*
no journal, ,
no abstracts in English
Nakayama, Shinsuke; Kono, Hiroshi*; Araki, Shohei*; Watanabe, Yukinobu*; Iwamoto, Osamu; Ye, T.*
no journal, ,
Theoretical model analysis of deuteron-induced activation cross sections were performed using the calculation code system we have developed so far. In our previous works, we analyzed mainly double differential cross sections for the 
and 
reactions, and activation cross sections from the 
reactions in order to validate calculation method for the direct processes. In the present work, we analyzed activation cross sections from multi particle emission induced by high energy deuteron in order to validate calculation method for statistical decay processes. In the result of analysis, it was found out the calculation method adopted in our code system is valid.
Utsuno, Yutaka; Yoshida, Sota*; Shimizu, Noritaka*; Otsuka, Takaharu*
no journal, ,
no abstracts in English
Sawamura, Masaru; Umemori, Kensei*; Sakai, Hiroshi*; Furuya, Takaaki*; Enami, Kazuhiro*; Egi, Masato*
no journal, ,
no abstracts in English
Tsuji, Tomoyuki; Sakamoto, Yoshiaki; Hoshino, Yuzuru; Suzuki, Yasuo*; Machida, Hiroshi*
no journal, ,
JAEA is planning a business for the disposal of low level radioactive wastes generated from research, industrial, and medical facilities. Because those wastes are generated from various facilities, it is important to develop reasonable confirmation methods based on the characteristics of radioactive wastes. As a model case of development of the evaluation method to determine the radioactivity concentration, the common method was studied to determine the radioactivity concentration of PIE wastes stored in NDC. The radioactivity concentrations of 17 nuclides (Sr-90, Tc-99, U-235, 238, Pu-238, 239+240, 241, Am-241, Cm-244 and so on) were calculated by ORIGEN-2 based on actual data such as initial contents and operation record of the spent fuel. From the comparison of the obtained data by radiological measurement with calculated values, it was studied that the theoretical method was applied to determine the radioactivity concentrations of 17 nuclides of PIE wastes.
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
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
no journal, ,
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
Jiao, L.; Takase, Kazuyuki; Liu, W.; Nagatake, Taku; Uesawa, Shinichiro; Yoshida, Hiroyuki; Shibata, Mitsuhiko
no journal, ,
To construct a database for upwards air/water flows in a vertical pipe, extensive measurements of air/water flows in a vertical pipe using the wire-mesh sensor technology were conducted at the thermal fluid dynamic test facility TPTF of the Japan Atomic Energy Agency. The test section is 4m in length and 58mm in inner diameter, two sets of three-layers-WMS were set separately at the 1.15m and 1.65m elevation of the air injection position. Air was injected from the bottom of the pipe through 0.6mm/1mm/2mm diameter nozzles. The obtained data are characterized particularly by their quantity and their detailed information on important two-phase flow parameters (e.g. radial distribution of the void fraction, the gas velocity and the time and cross-section averaged bubble size distribution for different test section heights). In the near future, we would like to use the WMS to measure the detailed two-phase flow in sub-channels of a simulated bundle flow.
Nagaya, Yasunobu
no journal, ,
no abstracts in English