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Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, X.; Kunitomi, Kazuhiko
JAERI-Tech 2002-087, 83 Pages, 2002/11
no abstracts in English
Sugino, Kazuteru; Iwai, Takehiko*;
JNC TN9400 2000-098, 182 Pages, 2000/07
In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 assembly, which is the first core of the BFS-62 series. The core contains the enriched U0 fuel surrounded by the U0 blanket. The standard analytical method for fast reactors has been applied, which was used for the JUPITER and other experimental analyses. Due to the lack of the analytical data the 2D RZ core calculation was mainly used. The 3D XYZ core calculation was applied only for the preliminary evaluation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality and the reaction rate ratio. However, it was found that accurate evaluation of the reaction rate distribution was impossible without exact consideration of the arrangement of the two types of sodium (with and without hydrogen impurity), which can be accommodated by the 3D core analysis, thus it was essentia1. In addition, it was clarifie that there was a room for an improvement of the result on the reaction rate distribution in the blanket and shielding regions. The application of the 3D core calculation improved the result on the control rod worth because 3D core model can more exactly consider the shape of the control rod. Furthermore it was judged that the result of the analysis on the sodium void reactivity .....
; *; *; *
JNC TN8410 2000-011, 185 Pages, 2000/05
This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.
Akie, Hiroshi; Takano, Hideki; Anoda, Yoshinari; Muromura, Tadasumi
Proc. of Int. Conf. on Future Nuclear Systems (Global'97), 2, p.1136 - 1141, 1997/00
no abstracts in English
Nakano, Yoshihiro
JAERI-Tech 95-002, 63 Pages, 1995/02
no abstracts in English
*
PNC TN9410 92-278, 347 Pages, 1992/09
A series of critical experiments for conventional two-zone homogeneous cores of 6 to 8 MWe-class LMFBR, JUPITER phase-I, were analyzed and evaluated using the latest analytical method, which had been established from the preceding numerous studies on fast reactor physics. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The analytical method and results are summarized as follows: (1)Analytical method (a)Nuclear data : 70-group fast reactor constant set JFS-3-J2(1989 edition) based on the Japanese Evaluated Nudear Data Library, version 2 (JENDL-2). (b)Cell calculation : plate stretch model, cell heterogeneity treatment by Tone's method and transport cross-sections weighted with neutron current. (c)Base core calculation : 18-group, three-dimensiona1 XYZ diffusion theory and Benoist's anisotropic diffusion coefficients. (d)Correction calculation : three-dimensional transport effect, mesh size effect, cell asymmetric effect and all master model effect etc. (2)Analytical results (a)The C.E (calculation/experiment) values of criticality agree quite well among seven cores (ZPPR-910D/2) and do not depend on the core volume or the number of control rod positions (CRP). (b)The C/E values of control rod worths increase gradually from the core center to the core edge positions in each core (511%). Those of reaction rate distributions also indicate similar spatial variations (25%), which is considered to be consistent with the C/E tendency of control rod worths. (c)The reaction rate ratios of C28/F49 and F25/49 give quite stable C/E values of 1.06 and 1.03, respectively. (d)The C/E values of sodium void reactivities are overestimated by +25% at core center region. The C/E dependence on void region size, which was pointed out in the past analyses, is found in the ZPPR-9 core, but not in the ZPPR-10 series. (e)The C/E dependence of 4% on the radial positions were found in sample ...
; Iijima, Tsutomu; ; ; ; ; ; *; ;
JAERI-M 9057, 25 Pages, 1980/09
no abstracts in English
; *; ; Iijima, Tsutomu; *; ;
JAERI-M 9055, 63 Pages, 1980/09
no abstracts in English
JAERI-M 7753, 176 Pages, 1978/08
no abstracts in English
JAERI-M 7135, 30 Pages, 1977/06
no abstracts in English
Maeda, Shigetaka; Itagaki, Wataru; Maeda, Koji; Maki, Ryosuke*; Yoshida, Katsumi*
no journal, ,
Highly microstructure-controlled ceramic neutron absorbers were developed for improving safety of fast reactors. The control rod worth was evaluated using several compositions as parameter. It was confirmed that the effect of additives on control rod worth was negligible.