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Kamei, Miho; Ozawa, Takayuki
no journal, ,
The future fast reactor (FR) cycle system to reduce radioactive waste volume by recycling Pu and minor actinides (MA), and it is important to evaluate the influence of high Pu content and MA content on oxide fuel performance. In this study, we investigated the influence of fuel characteristics on power-to-melt (PTM) from a viewpoint to prevent fuel melting. As a result, PTM of FR oxide fuel with high Pu content and MA content would be affected by fuel properties due to fuel composition, but it revealed that a correlation between fuel specifications and PTM in MA-containing MOX fuel would be similar to that in MOX fuel, and it can be mentioned that thermal performance of MA-containing MOX fuel would be secured to be equal to that of conventional MOX fuels by providing suitable fuel specifications up to MA content of 5 wt.%.
Ozu, Akira; Takase, Misao*; Haruyama, Mitsuo; Kurata, Noritaka*; Kobayashi, Nozomi*; Tobita, Hiroshi; Kureta, Masatoshi; Nakamura, Tatsuya; To, Kentaro; Suzuki, Hiroyuki; et al.
no journal, ,
no abstracts in English
Shibata, Keiichi
no journal, ,
Although JENDL-4.0 was released in 2010, the high-energy cross sections of about 40 FP nuclides were not re-examined due to the time limit. Tellurium is one of such data which were not revised in the high-energy region. Considering the significance of the activation cross sections of tellurium, the present work was undertaken to evaluate the neutron cross sections of 
Te in the energy from 10
eV to 20 MeV. The statistical model code CCONE, which was based on the Hauser-Feshbach theory, was used to evaluate cross sections above 10 eV. Parameters required as input to the CCONE code were determined by reflecting the latest knowledge on nuclear physics. As for the neutron optical model parameters, we employed the values obtained by Kunieda et al. using the coupled-channel method. The evaluated results reproduce existing experimental data very well.
Nagai, Takayuki; Okamoto, Yoshihiro; Kano, Shigeru; Nishizawa, Daiji; Seki, Katsumi*; Homma, Masanobu*; Kobayashi, Hiromi*; Ayame, Yasuo
no journal, ,
no abstracts in English
Tc and 
I in the keV regionIwamoto, Nobuyuki
no journal, ,
no abstracts in English
Inose, Takehiko*; Nishizawa, Daiji; Oyama, Koichi; Miyauchi, Atsushi; Nagai, Takayuki
no journal, ,
no abstracts in English
Kimura, Atsushi; Nakamura, Shoji; Terada, Kazushi; Nakao, Taro; Igashira, Masayuki*; Katabuchi, Tatsuya*; Hori, Junichi*
no journal, ,
no abstracts in English
Iwamoto, Hiroki; Nishihara, Kenji; Yagi, Takahiro*; Pyeon, C. H.*
no journal, ,
no abstracts in English
)Terada, Kazushi; Nakamura, Shoji; Kimura, Atsushi; Nakao, Taro; Harada, Hideo; Igashira, Masayuki*; Katabuchi, Tatsuya*; Hori, Junichi*
no journal, ,
no abstracts in English
Nakao, Taro; Kimura, Atsushi; Nakamura, Shoji; Terada, Kazushi; Harada, Hideo; Igashira, Masayuki*; Katabuchi, Tatsuya*; Hori, Junichi*
no journal, ,
The absolute amount of the test sample is required for the neutron capturing cross-section measurements. But the absolute amount has not been precisely determined for obtained samples. Therefore, it is necessary to determine absolute amount of samples accurately by nondestructively. This presentation will report for the future plan about the all deposit heat measurement from minor actinides samples in order to determine the absolute amount of samples non-destructively. This presentation is as part of nuclear system research and development project "R&D for accuracy improvement of neutron nuclear data on minor actinides".
Nakamura, Shoji; Kimura, Atsushi; Terada, Kazushi; Nakao, Taro; Harada, Hideo; Igashira, Masayuki*; Katabuchi, Tatsuya*; Hori, Junichi*; Fujii, Toshiyuki*
no journal, ,
As a part of "R&D for accuracy improvement of neutron nuclear data on minor actinides", technical developments for accurate determination of thermal-neutron capture cross-sections have been promoted. In this presentation, the experimental plan, the present situation of the research and obtained results will be presented.
Hamada, Hirotsugu; Uchibori, Akihiro; Ohshima, Hiroyuki
no journal, ,
For the purpose of a safety evaluation of heat transfer tube failure in the FBR steam generator, the long-term leak enlargement and propagation analysis code (LEAP-III) is under development. A model of overheating tube rupture was incorporated into LEAP-III and LEAP-III was applied to an analysis of SWAT-3 test to evaluate the applicability of the code.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Mizuyama, Kazuhito
no journal, ,
Under Innovative Nuclear Research and Development Program, "R&D for accuracy improvement of neutron nuclear data on minor actinides", present status of measurements and evaluations for MA nuclear data has been reviewed for "High quality evaluation via bouncing ideas off measurement and evaluation". The important MAs of Np-237 and Am-241, 243 were investigated in the thermal to fast neutron region. It was found that measured data for the thermal neutron cross sections of Np-237 and Am-241 were converging to a value with 5 to 6% discrepancies. The errors in evaluated nuclear data library were checked by comparing with the experimental data. Data for important FPs of Tc-99, I-129 etc. were also reviewed. In this presentation, a plan and result of review for a part of nuclear data evaluation of the program will be reported.
Harada, Hideo; Iwamoto, Osamu; Nakamura, Shoji; Kimura, Atsushi; Iwamoto, Nobuyuki; Terada, Kazushi; Nakao, Taro; Mizuyama, Kazuhito; Igashira, Masayuki*; Katabuchi, Tatsuya*; et al.
no journal, ,
The research project has been started for improving accuracy of neutron nuclear data for minor actinides (MAs) and long-lived fission products (LLFPs), which is required for developing innovative nuclear system transmuting these nuclei. The project consists of 5 items: (1) Accurate measurements of thermal neutron capture cross-sections (2) High-precision quantification of sample amount used for TOF measurement (3) Resonance parameter determination by combining total and capture cross sections (4) Extension of capture cross sections to high energy neutrons (5) High quality evaluation based on iterative communication with experimenters. The overall plan of the project is presented.
Yamashita, Susumu; Yoshida, Hiroyuki; Takase, Kazuyuki
no journal, ,
In the Fukushima accidents, fuel assemblies which were installed in the reactors were reached a high temperature by stop of the core cooling system with a power station black-out. As a result, it is considered that the core degradation has been introduced because the melting of the fuel rods occurred and the melting behavior was expanded. In order to elucidate a progress of the melting phenomena in the reactor core, a numerical simulation code which can be precisely evaluated the melting phenomena is required. Therefore, we have been developing the numerical code named JUPITER to elucidate such phenomena. In the last report, we implemented multicomponent analysis function to the JUPITER and confirmed that it works well for the function in molten material relocation analysis including a heating and non heating material. In this study, we will present the result of comparison between the corium spreading experiment and its numerical result.
Suzuki, Kazuhiro; Toyokawa, Takuya; Motooka, Takafumi; Tsukada, Takashi; Ueno, Fumiyoshi; Terakawa, Yuto; Suzuki, Miho; Ichise, Kenichi; Numata, Masami; Kikuchi, Hiroyuki
no journal, ,
no abstracts in English
Okagaki, Yuria; Shibamoto, Yasuteru; Sun, Haomin; Satou, Akira; Yonomoto, Taisuke
no journal, ,
no abstracts in English
Amamoto, Ippei; Kobayashi, Hidekazu; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
no journal, ,
The immobilization method by iron phosphate glass (IPG) medium is one of the candidate techniques for manufacturing waste forms of sludge arising from the treatment of contaminated water at the stricken Fukushima Dai-ichi Nuclear Power Plant. In this paper, some thermodynamic values for the theoretical analysis of vitrification were estimated to make up the calculated phase diagrams. These calculated phase diagrams were then compared with experimental results.
Yokoyama, Kaoru; Ohara, Yoshiyuki; Sugitsue, Noritake; Takahashi, Nobuo; Rong, D.*; Takeda, Hiroshi*; Kochi, Toshinori*; Yanase, Shinichiro*; Kuwagi, Kenya*; Takami, Toshihiro*; et al.
no journal, ,
Some disaster wastes polluted with the radioactive cesium diffused in the accident of the Fukushima Daiichi Nuclear Power Plant are incinerated in the refuse incineration plant. It is thought that cesium is not emitted outside since cesium is coagulated to the refuse incineration ash and trapped by a bag filter or an electric dust collector. Therefore, we report the simulation of the behavior of cesium in an incineration plant.
Kinase, Sakae; Sato, Satoshi; Takahashi, Tomoyuki*; Saito, Kimiaki
no journal, ,
no abstracts in English