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Sugino, Kazuteru; Iwai, Takehiko*;
JNC TN9400 2000-098, 182 Pages, 2000/07
JNC-TN9400-2000-098.pdf:5.74MB
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 .....
Hunter
JNC TN9400 99-049, 74 Pages, 1999/04
This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2d) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of keff was 1.1%
k/k higher than the measured value, Na void worth C/E values were 
1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes, though the efect should be investigated in any future experiments.) several sample worth values were small compared with calculational uncertaint
; Takano, Hideki; ; A.G.Morozov*; V.S.Smirnov*; V.V.Orlov*
Proc. of ARS94 Int. Topical Meeting on Advanced Reactors Safety,Vol. 1, 0, p.544 - 548, 1994/00
no abstracts in English
Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki; Usami, Shin; Takeda, Toshikazu*
no journal, ,
no abstracts in English