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Ikusawa, Yoshihisa; Nagayama, Masahiro*
JAEA-Data/Code 2023-006, 24 Pages, 2023/07
JAEA-Data-Code-2023-006.pdf:1.42MB
Core fuels with stainless steel cladding and high plutonium content mixed oxide (MOX) fuel in a water-cooled environment, such as supercritical water-cooled reactors (SCWR) and reduced-moderation water reactors (RMWR), have been studied. In order to contribute to the research and development of such a core fuel concept, the fuel performance code "FEMAXI-8" was verified based on the results of post irradiation examinations of MOX fuel irradiated in the experimental fast reactor "JOYO". FEMAXI-8 is the latest version of the behavior analysis code developed by JAEA to analyze the behavior of light water reactor fuels under normal operation and transient conditions. This latest code has been improved and developed to allow the selection of stainless steel cladding property models to analyze improved fuels such as accident tolerant fuels. The purpose of this report is to confirm the prediction accuracy of FEMAXI-8 for the irradiation behavior of the new type of core fuel that is currently being developed. As a result of the verification, it was confirmed that FEMAXI-8 has sufficient analysis accuracy for the irradiation behavior of sodium-cooled fast reactor MOX fuel with stainless steel cladding, which exceeds the plutonium content and irradiation conditions of light water reactors. In the future, the analysis accuracy of FEMAXI-8 could be improved by adopting the O/M ratio dependence of MOX fuel thermal conductivity and the irradiation behavior evaluation model at high temperature.
Ogata, Takanari*; Takano, Masahide
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(7), p.541 - 546, 2021/07
This is a commentary on metallic fuels for fast reactors and nitride fuels for minor actinide transmutation in accelerator driven system, as the 4th article of serial lecture on Journal of the Atomic Energy Society of Japan; Concepts and basic designs of various nuclear fuels.
Kihara, Yoshiyuki; Tanaka, Kosuke; Koyama, Shinichi; Yoshimochi, Hiroshi; Seki, Takayuki; Katsuyama, Kozo
NEA/NSC/R(2017)3, p.341 - 350, 2017/11
In order to investigate the effect of the addition of americium to MOX fuels on the irradiation behaviour, the "Am-1" program is being conducted at the JAEA. The Am-1 program consists of two short-term irradiation tests of 10-min and 24-h irradiation periods, and a steady-state irradiation test. The short-term irradiation tests and their post irradiation examinations (PIEs) have been successfully completed. To date, the data for PIE of the Am-MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation have been obtained and reported. In this paper, the results obtained from the Am-1 program are reviewed and detailed descriptions of the fabrication and inspection techniques for the Am-MOX fuels prepared for the program are provided. PIE data for the Am-MOX fuels at the initial stage of irradiation have been accumulated. In this paper, unpublished PIE data for the Am-MOX fuels are also presented.
Ozawa, Takayuki; Ikusawa, Yoshihisa; Kato, Masato
Transactions of the American Nuclear Society, 113(1), p.622 - 624, 2015/10
A recycle system for minor actinides (MAs), in which MAs are recycled by reprocessing and irradiating them in a fast reactor, is studied to reduce the degree of hazard and the amount of high-level radioactive wastes. MAs would be used as mixed oxide (MOX) fuels with plutonium and uranium in fast reactors. Since MA content of MA-bearing MOX (MA-MOX) to be used in fast reactors is assumed to reach 
5 wt%HM, the effects on not only fuel properties but also fuel behaviors have to be estimated to use MA-MOX as fast reactor fuels. As the MOX fuels to be used will be irradiated at a comparably high linear power and the fuel center temperature would be assumed to be over 2,273 K during irradiation in the fast reactors, fuel restructuring would take place due to void migration towards the fuel center under the radial temperature gradient, and a central void would be formed. Since the fuel center temperature would be decreased by the effect of formation of the central void, the fuel restructuring is one of the most important behaviors for fast reactor fuels. In this study, the effect of MA content on fuel restructuring behavior was estimated from the results of irradiation experiments such as B11 and B14 performed in Joyo to study the irradiation behaviors of MA-MOX and the calculation results using a fuel restructuring model which can take into account MA-MOX dependence on vapor pressure.
Committee of the Halden Joint Research Programme
JAERI-Tech 2004-023, 38 Pages, 2004/03
JAERI-Tech-2004-023.pdf:1.85MB
JAERI has performed cooperative researches with several Japanese organizations utilizing the Halden Boiling Heavy Water Reactor(HBWR) which is located at Halden in Norway. These researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summarizes the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 2000 January to 2002 December. During the period, seven cooperative researches had been carried out. Two of them had been completed and other five researches have been continued to the next three years period. Most of them are irradiation test researches of advanced fuel and cladding in order to prepare the higher burnup utilization and introduction of LWR fuel and MOX fuel in LWRs of Japan.
Yamashita, Toshiyuki; Akie, Hiroshi; Nakano, Yoshihiro; Kuramoto, Kenichi; Nitani, Noriko; Nakamura, Takehiko
Progress in Nuclear Energy, 38(3-4), p.327 - 330, 2001/02
Times Cited Count：12 Percentile：64.73(Nuclear Science & Technology)no abstracts in English
; ;
JNC TN9400 2000-041, 29 Pages, 2000/03
JNC-TN9400-2000-041.pdf:1.18MB
Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.
Harada, Katsuya; Nishino, Yasuharu; Mita, Naoaki; Amano, Hidetoshi
JAERI-Tech 2000-031, p.27 - 0, 2000/03
JAERI-Tech-2000-031.pdf:2.95MB
no abstracts in English
Yamashita, Toshiyuki; Akie, Hiroshi; Kimura, Hideo; Takano, Hideki; Muromura, Tadasumi
IAEA-TECDOC-1122, p.309 - 320, 1999/11
no abstracts in English
; Mita, Naoaki; Nishino, Yasuharu; Amano, Hidetoshi
JAERI-Conf 99-009, p.103 - 111, 1999/09
no abstracts in English
Saito, Hioraki*; Iriya, Yoshikazu*
JNC TJ8440 99-003, 156 Pages, 1999/03
no abstracts in English
Shiba, Kiyoyuki; Hishinuma, Akimichi
Purazuma, Kaku Yugo Gakkai-Shi, 74(5), p.436 - 441, 1998/05
no abstracts in English
Noda, Kenji
JAERI-Conf 98-006, 286 Pages, 1998/03
no abstracts in English
Suzuki, Yasufumi; Ogawa, Toru; Arai, Yasuo; Mukaiyama, Takehiko
Actinide and Fission Product Partitioning and Transmutation, p.213 - 221, 1998/00
no abstracts in English
Suzuki, Yasufumi; Arai, Yasuo
Purutoniumu Nenryo Kogaku; Nihon Genshiryoku Gakkai "Jisedai Nenryo" Kenkyu Semmon Iinkai, p.260 - 291, 1998/00
no abstracts in English
Suzuki, Yasufumi; Arai, Yasuo; Iwai, Takashi; Nakajima, Kunihisa
Proc. of Int. Conf. on Future Nuclear Systems (Global'97), 1, p.522 - 527, 1997/00
no abstracts in English
; ; ; Oyamada, Rokuro; Saito, Minoru
Proc. of 4th Asian Symp. on Research Reactors, 10 Pages, 1993/00
no abstracts in English
; H.Cords*; H.Nickel*
Journal of Nuclear Science and Technology, 29(9), p.851 - 858, 1992/09
no abstracts in English
