The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8Â8-2 UO 2 assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of 235 U, well reproduce those of 239 Pu and 240 Pu, yet underestimate those of 236 U, 237 Nd, 238 Pu, 241 Pu, and 242 Pu. The inventories of minor actinides are underestimated by the present analysis except for 241 Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8Â8-4 UO 2 assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of AE10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8Â8-4 UO 2 assemblies were well reproduced by the burnup calculations of MVP-BURN.
A few-group nodal BWR core simulator NEREUS, which is based on the analytic polynomial nodal method, has been developed. The analytic polynomial nodal method is applicable to multigroups and has good accuracy for analysis of cores having large spectral mismatch between fuel assemblies, since the intranodal thermal fiux can be correctly represented. In NEREUS, the following advanced methods are adopted to overcome the shortcomings of the conventional analytic polynomial nodal method. The flux solution iteration matrix is cast into the form of finite difference, by using the flux discontinuity factors which correct the finite difference error as well as the homogenization error. The intranodal burnup and spectral history distributions are considered in the method, and source moments are obtained by orthogonal expansions. Burnup calculations are made by looking-up exposure dependent tables for macroscopic cross sections, which are prepared by single assembly spectrum and depletion calculations. A unified model accounting for effects of spectral histories, caused by the spectral interaction between fuel assemblies and the control blade insertion, was incorporated. The NEREUS code was verified against benchmark problems and core tracking calculations.
Fuel rod gamma-ray spectrometry was performed for one-, two-, three-, and five-cycle irradiated BWR 8 Â 8-4 fuel assembles, and relative rod-by-rod FP inventory distributions of 137 Cs, 134 Cs, 106 Ru, and 95 Zr were obtained for the upper and lower axial height nodes of the assemblies. The measured data was corrected for the difference in gamma-ray transmission between UO 2 and Gd 2 O 3 -UO 2 rods. These distributions were compared with those calculated using the collision probability method, Pij, of the SRAC code system with an infinite lattice model of the fuel assembly. The calculated results generally well reproduce the measured distributions, and the accuracy of the analysis method of the present study was evaluated to be 1 to 2% for the inventory distributions of 137 Cs, 2 to 3% for 134 Cs, 2 to 3% for 106 Ru, and 2 to 3% for 95 Zr, which represent the distributions of the burn-up, the thermal flux multiplied by the burn-up, the buildup of 239 Pu, and the fission rate at the end of the fuel discharge cycle, respectively. One of the notable results of the analysis of this study is that the FP inventories for the Gd 2 O 3 -UO 2 rods were underestimated in most cases.
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