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Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*
JAEA-Research 2024-012, 98 Pages, 2025/02
JAEA-Research-2024-012.pdf:32.24MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO
(PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.
pellet in molten Zr claddingIto, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*
Journal of Nuclear Science and Technology, 60(4), p.450 - 459, 2023/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Udagawa, Yutaka; Tasaki, Yudai
JAEA-Data/Code 2021-007, 56 Pages, 2021/07
JAEA-Data-Code-2021-007.pdf:5.05MB
Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.
Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio
JAEA-Data/Code 2019-018, 22 Pages, 2020/01
JAEA-Data-Code-2019-018.pdf:1.39MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.
Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06
Times Cited Count：12 Percentile：69.57(Nuclear Science & Technology)no abstracts in English
Mori, Takamasa; Kojima, Kensuke*; Suyama, Kenya
JAEA-Research 2018-010, 57 Pages, 2019/02
JAEA-Research-2018-010.pdf:6.25MB
In order to estimate applicability of the statistical geometry model (STGM) of MVP/GMVP, a parametric study in infinite geometry and criticality safety analyses for direct disposal of spent fuel in simple finite geometry have been carried out by using the MVP Monte Carlo code. It has been found that calculations with STGM for larger fuel spheres give larger thermal utilization factors and larger infinite multiplication factors compared with explicit random models in the range of fuel sphere packing fraction between 6.5 % and 63.3 %. Substantial differences are not observed between the results with two nearest neighbor distributions (NNDs); that given by the MCRDF code and the analytical expression based on a statistically uniform distribution. It is inferred that the overestimation by STGM is caused by the facts that STGM cannot take account of the surroundings of each neutron, whether a fuel sphere rich region or a water moderator rich one, because STGM always uses an NND averaged over such surroundings and that STGM, therefore, cannot take the effect of consecutive scatterings in the water moderator into account.
Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki
JAEA-Data/Code 2018-016, 79 Pages, 2019/01
JAEA-Data-Code-2018-016.pdf:2.75MB
FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.
Onozawa, Atsushi; Kushida, Teruo; Kanazawa, Hiroyuki
JAERI-Tech 2004-061, 39 Pages, 2004/11
JAERI-Tech-2004-061.pdf:8.64MB
The swelling observed on irradiated fuels is caused by the accumulation of fission products and irradiation defects. The swelling ratio is changed along with radius region in the pellet due to burn up difference caused by that of neutron flux. To investigate the swelling behavior at the small area of the pellet, it is needed to measure the density of fuel fragments picked from an irradiated pellet. In this circumstance, once-through type densitometer was developed to measure the density of the small irradiated specimen precisely and to handle the samples easily with remote control systems. Several kinds of metallic and ceramic standard specimens are prepared to investigate the dependence of the sample weight, density and porosity on the accuracy. The results of characteristic examination using these specimens indicate that this densitometer has enough accuracy. In addition, some parts of this apparatus are controlled by motor drive units, which made it possible to measure the density full-automatically.
Nakamura, Jinichi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Kanazawa, Toru; Sasajima, Hideo
JAERI-Tech 2003-008, 32 Pages, 2003/03
JAERI-Tech-2003-008.pdf:1.49MB
no abstracts in English
Ando, Yoshihira*; Nishihara, Kenji; Takano, Hideki
Journal of Nuclear Science and Technology, 37(10), p.924 - 933, 2000/10
no abstracts in English
fuel discsUne, Katsumi*; Nogita, Kazuhiro*; Shiratori, Tetsuo; Hayashi, Kimio
Journal of Nuclear Materials, 288(1), p.20 - 28, 2000/09
Times Cited Count：20 Percentile：78.30(Materials Science, Multidisciplinary)no abstracts in English
fuel under reactivity initiated accident conditionsSasajima, Hideo; Nakamura, Jinichi; Fuketa, Toyoshi; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 36(11), p.1101 - 1104, 1999/11
Times Cited Count：2 Percentile：21.18(Nuclear Science & Technology)no abstracts in English
Yanagisawa, Kazuaki; Omichi, Toshihiko*; Shirasu, Noriko; Muromura, Tadasumi;
JAERI-Tech 99-032, 65 Pages, 1999/03
no abstracts in English
pelletNakamura, Jinichi; Uchida, Masaaki; Uetsuka, Hiroshi; Furuta, Teruo
IAEA-TECDOC-1036, 0, p.127 - 138, 1998/08
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
JAERI-Data/Code 97-046, 210 Pages, 1997/11
JAERI-Data-Code-97-046.pdf:5.41MB
no abstracts in English
pelletNakamura, Jinichi; Uchida, Masaaki; Uetsuka, Hiroshi; Kodaira, Tsuneo; Yamahara, Takeshi;
Proc. of Int. Topical Meeting on LWR Fuel Performance, 0, p.499 - 506, 1997/03
no abstracts in English
latticesNakajima, Ken;
Nuclear Technology, 113, p.375 - 379, 1996/03
Times Cited Count：5 Percentile：44.68(Nuclear Science & Technology)no abstracts in English
pellet irradiated at HBWRNakamura, Jinichi; Furuta, Teruo; Sukegawa, T
HPR-347, 12 Pages, 1996/00
no abstracts in English
Nakamura, Jinichi; Uetsuka, Hiroshi; Kono, Nobuaki; ; ; Furuta, Teruo
HPR-345 (Vol. II), 0, 13 Pages, 1995/00
no abstracts in English
fuel particlesMinato, Kazuo; Ogawa, Toru; Fukuda, Kosaku; ; ;
Journal of Nuclear Materials, 208, p.266 - 281, 1994/00
Times Cited Count：45 Percentile：94.83(Materials Science, Multidisciplinary)no abstracts in English
