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

Abstract

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.

Analysis of core physics experiments on irradiated BWR MOX fuel in the REBUS program

“…The measured inventories provided the burnup of the measured fuel sample by applying Neodymium-148 Method [12]. Ando et al reported the burnup value of the sample as 69.90 GWd/t [13]. They used the measured isotopic inventories of actinide nuclides and 148 Nd, and the effective values of MeV/fission and fission yield of 148 Nd which were calculated by a continuous energy Monte Carlo burnup calculation code MVP-BURN [14,15].…”

“…This value for 241 Am seems to be unreasonably high. In the MALIBU program, where two or three different organizations including SCK/ CEN participated in chemical assay of fuel samples, the other organization reported 10% and 17% larger measurement data of the total Am weight for the samples taken from the axially-adjacent parts in the fuel rod [13]. The other prominent C/Es are seen for some of the metallic FP nuclides, 99 Tc, 101 Ru, 108 Pd, and 109 Ag, which cause considerable contributions due to the C/Es of 1.267, 1.424, 1.300, and 1.260, respectively.…”

“…Since the burnup used in the inventory calculation of the sample was 5.3% larger than that of the destructive measurement, the inventories of FP nuclides, most of which lineally increase with burnup, were overestimated by the inventory calculation. The other trend is related to the infinite assembly model used in the inventory calculation, which caused overestimation of fissile nuclides such as 239 Pu, and 241 Pu due to neglecting neutron energy interaction from adjoining UO 2 assemblies during irradiation [3,13]. The other trend is seen in Am and metallic FP isotopes such as Ru and Ag.…”

“…This would be possibly related to measurement errors in the chemical assay technique, which has been discussed in Refs. [6,13].…”

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2fuels in the MALIBU program

“…As part of the above-mentioned validation study (i), the Japan Nuclear Energy Safety Organization has been participating in the international program MALIBU [2] on the measurement of isotopic compositions in high-burnup spent fuels discharged from light water reactors and performing neutronic analyses of the measurement data [3].…”