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Ozawa, Takayuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
In a recycle system for minor actinides (MAs) currently studied to reduce the degree of hazard and the amount of high-level radioactive wastes, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. MA content is expected to be 
5 wt.% in the future recycle system, and MAs might affect irradiation behavior of MA-MOX fuels. The main influences of MA-containing would be increase of fuel temperature and cladding stress, and the important behaviors would be fuel restructuring, redistribution, helium (He) generation and cladding corrosion. The MA-containing influences were evaluated with CEPTAR.V2, including fuel properties and analysis models to evaluate the MA-MOX fuel irradiation behavior, by using the results of highly americium (Am) containing MOX irradiation experiment, B8-HAM, performed in Joyo. The irradiation behavior of Am-MOX fuels could be precisely analyzed and revealed the influences of Am-containing.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi
HPR-362, Vol.2, 12 Pages, 2004/05
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.
Yamanaka, Shinsuke*; Abe, Kazuyuki
JNC TY9400 2000-004, 78 Pages, 2000/03
JNC-TY9400-2000-004.pdf:2.39MB
no abstracts in English
; *;
JNC TN9400 2000-045, 64 Pages, 2000/03
JNC-TN9400-2000-045.pdf:2.47MB
During the irradiation, the Pu redistribution phenomena would occur in the FBR MOX fuel pellets. The phenomena would considerably affect on the thermal properties of the fuels, therefore, it is need to establish the evaluation method for Pu redistribution phenomena. ln JNC, the efforts for development of the evaluation model for the phenomena had been continued and the simple evaluation model was constructed in 1992. In this work, the modification of the simple model developed in JNC has been done and the following results were obtained. (1)Based on the recent data of the MOX fuel irradiation tests, the evaluation model for Pu redistribution phenomena constructed in l992 is modified. And the model is included into the fuel performance analysis code "CEDAR". (2)To calibrate the modified CEDAR code, it is confirmed that the uncertainty in the Pu concentration evaluation for the center of the fuel pellet at EOL is about 
3wt.%. (3)Based on the results of the evaluations using the modified CEDAR code, it is found that, in the early stage of the irradiation, the Pu redistribution is controlled by the vapor transportation mechanism via pores, and after that, the Pu redistribution is kept in progress due to the thermal diffusion mechanism with the change of the Pu concentration due to the degradation of U and Pu by fissions. And it is also found that the O/M ratio dependence of the U-Pu inter diffusion coefficients would affect on the Pu redistribution mechanisms, in especial, in the early stage of the irradiation.
Yoshinaga, Makio; Nakamura, Takehiko; Yamazaki, Toshi*
JAERI-Tech 2000-017, p.59 - 0, 2000/03
JAERI-Tech-2000-017.pdf:2.31MB
no abstracts in English
Saito, Hioraki*; Iriya, Yoshikazu*
JNC TJ8440 99-003, 156 Pages, 1999/03
no abstracts in English
Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kikuchi, Keiichi*; Abe, Tomoyuki*
Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7) (CD-ROM), 10 Pages, 1999/00
no abstracts in English
*; Ishijima, Kiyomi; Yamahara, Takeshi
JAERI-Data/Code 98-002, 24 Pages, 1998/02
JAERI-Data-Code-98-002.pdf:1.02MB
no abstracts in English
Nakamura, Jinichi
Kaku Nenryo, (29), P. 15, 1998/00
no abstracts in English
Katanishi, Shoji; Ishijima, Kiyomi; ; Kikuchi, Teruo;
JAERI-Research 94-039, 54 Pages, 1994/11
JAERI-Research-94-039.pdf:5.2MB
no abstracts in English
Nakamura, Jinichi; *; Furuta, Teruo; *
JAERI-M 91-027, 36 Pages, 1991/03
no abstracts in English
Tanzawa, Sadamitsu; *; ; *; *
JAERI-M 90-232, 30 Pages, 1991/01
no abstracts in English
Soda, Kunihisa
Shimyureshon, 9(2), p.79 - 86, 1990/00
no abstracts in English
Ichikawa, Michio; ; Kawasaki, Satoru
Journal of Nuclear Science and Technology, 26(1), p.118 - 125, 1989/01
no abstracts in English
; *
Journal of Nuclear Science and Technology, 24(1), p.12 - 22, 1987/01
Times Cited Count：1 Percentile：19.35(Nuclear Science & Technology)no abstracts in English
; ; Hirano, Masashi;
Journal of Nuclear Science and Technology, 23(12), p.1107 - 1109, 1986/12
Times Cited Count：2 Percentile：42.61(Nuclear Science & Technology)no abstracts in English
Amaya, Masaki
no journal, ,
The objectives of the fuel safety research at JAEA are to evaluate the appropriateness of current Japanese regulatory criteria and their safety margins regarding light-water-reactor fuels, to provide the data for regulation for improved fuels which consist of cladding and fuel pellet with new materials, and to provide fuel behavior analysis codes which are applicable to regulatory activities. In this presentation, the outline of fuel safety research activities at JAEA is presented in addition to the progress in reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) experiments.
Maeda, Seiichiro
no journal, ,
The uranium-plutonium mixed oxide (MOX) fuel bearing several percent minor actinides is adopted as the fast reactor fuel to reduce the high level radioactive waste volume and its toxicities. Their physical properties such as melting points and thermal conductivities are investigated symmetrically. It is getting clear that the effect of MA bearing is not so significant. The MOX fuel pins including up to 5% of Am were irradiated in Joyo to confirm their irradiation performance. Some types of irradiation experiments using MA-MOX fuels are planned after Joyo restart. The obtained data will contribute to develop fuel pin performance codes for MA-MOX fuels. SmART (Small Amount of Reuse Fuel Test) cycle program is in progress to demonstrate MA recycles starting from FR spent fuels. Furthermore, long life cladding with ODS steel is under development to enhance the transmutation efficiency of MA in one cycle. Thus, R&Ds in FR systems advances steadily to solve the waste issue.
