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Takino, Kazuo; Oki, Shigeo
JAEA-Data/Code 2023-003, 26 Pages, 2023/05
Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.
Yamauchi, Michinori*; Hori, Junichi*; Ochiai, Kentaro; Sato, Satoshi; Nishitani, Takeo; Kawasaki, Hiromitsu*
Fusion Engineering and Design, 81(8-14), p.1577 - 1582, 2006/02
Times Cited Count:1 Percentile:9.94(Nuclear Science & Technology)no abstracts in English
Maekawa, Fujio; Meigo, Shinichiro; Kasugai, Yoshimi; Takada, Hiroshi; Ino, Takashi*; Sato, Setsuo*; Jerde, E.*; Glasgow, D.*; Niita, Koji*; Nakashima, Hiroshi; et al.
Nuclear Science and Engineering, 150(1), p.99 - 108, 2005/05
Times Cited Count:6 Percentile:40.41(Nuclear Science & Technology)A neutronic benchmark experiment on a simulated spallation neutron target assembly with 1.94, 12 and 24 GeV proton beams conducted by using the AGS accelerator at BNL/US was analyzed to investigate validity of neutronics calculations on proton accelerator driven spallation neutron sources. Monte Carlo particle transport calculation codes NMTC/JAM, MCNPX and MCNP-4A with associated cross section data in JENDL and LA-150 were used for the analysis. As a result, although the overall energy range was encompassed from GeV to meV, i.e., more than 12 orders of magnitude, calculated fast and thermal neutron fluxes agreed approximately within 40 % with the experiments. Accordingly, it was concluded that neutronics calculations with these codes and cross section data were adequate for estimating nuclear properties in spallation neutron sources.
Subcommittee on Improvement of Reactor Thermal-Hydraulic Analysis Codes
JAERI-Review 2000-002, p.105 - 0, 2000/03
no abstracts in English
Maekawa, Hiroshi; M.A.Abdou*
Fusion Engineering and Design, 28, p.479 - 491, 1995/00
no abstracts in English
Takatsu, Hideyuki; Mori, Seiji*; Yoshida, Hiroshi; Hashimoto, T.*; Kurasawa, Toshimasa; Koizumi, Koichi; Enoeda, Mikio; Sato, Satoshi; Kuroda, Toshimasa*; *; et al.
Fusion Technology 1992, p.1504 - 1508, 1993/00
no abstracts in English
Nagaoka, Yoshiharu; Komukai, Bunsaku; ; Koike, Sumio; Saito, Minoru;
JAERI-M 92-098, 81 Pages, 1992/07
no abstracts in English
; ; Shinohara, Yoshikuni; ; Akino, Fujiyoshi; ; ; Ono, Akio; ;
JAERI-M 86-125, 240 Pages, 1986/08
no abstracts in English
JAERI-M 85-116, 259 Pages, 1985/08
no abstracts in English
Matsuura, Shojiro
JAERI-M 83-129, 246 Pages, 1983/09
no abstracts in English