Validation of a Continuous-Energy Monte Carlo Burn-up Code MVP-BURN and Its Application to Analysis of Post Irradiation Experiment

SUGOU³

2011

J. Nucl. Sci. Technol.

Analysis of measured isotopic compositions of four high-burnup BWR MOX fuel samples was performed by using a general-purpose neutronic calculation code SRAC and a continuous-energy Monte Carlo burnup code MVP-BURN. The initial Pu fissile content of the samples was 5.52 wt%, and the burnups ranged from 50 to 80 GWd/t. It is confirmed that a geometrical model including the effect of UO 2 assemblies adjacent to the MOX assembly is necessary in the burnup calculations to obtain accurate calculated isotopic compositions. The calculated results of MVP-BURN with JENDL-3.3 taking such effect into account show more accurate results for major actinides (U, Pu, and Am isotopes) and most fission products than those of infinite assembly calculations. The paper also shows the results calculated using SRAC with JENDL-3.3, ENDF/B-VII, and JEFF-3.1.

Section: Analysis Of Measured Isotopic Compositionsmentioning

confidence: 99%

Analysis of Measured Isotopic Compositions of High-Burnup BWR MOX Fuel

Ando

SUGOU³

2011

J. Nucl. Sci. Technol.

“…A Gd 2 O 3 -UO 2 pellet has 10 radial regions of equal volume. The temperatures of the pellets were postulated to be 627 C. The temperatures of the fuel cladding and the moderator were set to be 286 C. The burnup calculations for the interpolation analysis were also performed using a continuous-energy Monte Carlo burnup calculation code, MVP-BURN, 3) with the nuclear data library, JENDL-3.2 20) in order to inquire discrepancies between the analysis results of the SRAC calculations and those of the measurements for the 8Â8-4 assemblies, which is mentioned in the later section. The analysis conditions of the burnup calculations with MVP-BURN are similar to the SRAC calculations.…”

Section: Analyses Of Pellet Average Data (1) Analysis Methodsmentioning

confidence: 99%

“…Postirradiation examination data include valuable information that can be used to validate nuclear methodologies related to burnup calculations. One of them is isotopic inventories of irradiated fuels, and there are a large number of studies [2][3][4][5][6][7][8][9][10][11][12] on the nuclear analysis of the measurement data. The other field is the analysis 13,14) of rod-by-rod FP inventory distributions measured by fuel rod gamma scanning.…”

Section: Introductionmentioning

confidence: 99%

Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO₂Fuel Assemblies

Journal of Nuclear Science and Technology

Yamamoto²

2008

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. 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.

“…The pellet radial inventory distributions nðr i Þ of the FP nuclides were calculated by using MVP-BURN, 11) a continuous energy Monte Carlo code coupled with a burn-up calculation module. The calculation was performed in a two-dimensional geometry, and material conditions representing the irradiated BWR 8 Â 8-4 UO 2 fuel assembly are shown in Table 1.…”

Section: Gamma Spectrometrymentioning

confidence: 99%

Analysis of Rod-by-Rod FP Inventory Distributions in BWR 8×8 UO2 Assemblies Using Lattice Physics Method

Yamamoto²

2008

J. Nucl. Sci. Technol.

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.