Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo
Annals of Nuclear Energy, 162, p.108454_1 - 108454_7, 2021/11
Times Cited Count：1 Percentile：15.7(Nuclear Science & Technology)Okita, Shoichiro; Fukaya, Yuji; Goto, Minoru
Journal of Nuclear Science and Technology, 58(1), p.9 - 16, 2021/01
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Suppressing the kernel migration rates, which depend on both the fuel temperature and the fuel temperature gradient, under normal operation condition is quite important from the viewpoint of the fuel integrity for High Temperature Gas-cooled Reactors. The presence of the ideal axial power distribution to minimize the maximum kernel migration rate allows us to improve efficiency of design work. Therefore, we propose a new method based on Lagrange multiplier method in consideration of thermohydraulic design in order to obtain the ideal axial power distribution to minimize the maximum kernel migration rate. For one of the existing conceptual designs performed by JAEA, the maximum kernel migration rate for the power distribution to minimize the maximum kernel migration rate proposed in this study is lower by approximately 10% than that for the power distribution as a conventional design target to minimize the maximum fuel temperature.
Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo
Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08
Times Cited Count：1 Percentile：10.6(Nuclear Science & Technology)The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.
Nakatsuka, Toru; Tamai, Hidesada; Kureta, Masatoshi; Okubo, Tsutomu; Akimoto, Hajime; Iwamura, Takamichi
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 6 Pages, 2003/09
It is important to evaluate thermal margin of the tight lattice core in the Reduced-Moderation Water reactor (RMWR). In the present study, to assess the applicability of subchannel analysis for tight lattice cores, tight lattice CHF experiments were analyzed with COBRA-TF code. For the axial uniform heated rod bundle, the code gives good prediction of critical power for mass velocity of around 500kg/(m
s), while the code underestimates it for lower mass velocity and overestimates for higher mass velocity. The predicted BT position was outer channels and differed from the measured position. For the axially double-humped heated bundle, the code gives good prediction for mass velocity of around 200kg/(ms), and overestimates for higher mass velocity. It turned out that the two-phase multiplier of friction loss have a large influences on the flow distribution among the subchannels. To improve the calculation accuracy, it is required to predict precisely the flow distribution including the prediction of pressure distribution in a tight lattice bundle.
Obara, Toru*; Nakajima, Ken; *; Sekimoto, Hiroshi*; Suzaki, Takenori
JAERI-M 94-004, 40 Pages, 1994/02
no abstracts in English
C for HTTRYamashita, Kiyonobu; Maruyama, So; Murata, Isao; Shindo, Ryuichi; Fujimoto, Nozomu; Sudo, Yukio; Nakata, Tetsuo*; *
Journal of Nuclear Science and Technology, 29(5), p.472 - 481, 1992/05
no abstracts in English
C for High Temperature Engineering Test Reactor(HTTR)Yamashita, Kiyonobu; Maruyama, So; Murata, Isao; Shindo, Ryuichi; Fujimoto, Nozomu; Sudo, Yukio; Nakata, Tetsuo*; *
Proc. of the 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering,Vol. 1, p.419 - 424, 1991/00
no abstracts in English
;
Nuclear Technology, 78(9), p.207 - 215, 1987/09
no abstracts in English
; ; ;
JAERI-M 87-059, 57 Pages, 1987/05
no abstracts in English
Ueki, Taro
no journal, ,
A new methodology has been developed to make the reliable estimation of statistical errors in Monte Carlo criticality calculation (MCCC). The methodology developed is directly based on the convergence process in the functional central limit theorem and is shown to perform well in the evaluation of reactor power distribution. The theoretical backbones are described within the general context as framed in the operations research. The requisite basics of statistics are reviewed in terms of output analysis in MCCC. Numerical results are presented for the initial core model of a 1200 MWe pressurized water reactor. Preliminary results of fractal dimension analysis are shown to discuss a potential for convergence assessment.
