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

Abstract

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO 2 assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 6 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 6 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 6 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO 2 sample of 72 GWd/t.

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

“…Those shown as "PWR 15 × 15 UO2" and "PWR 15 × 15 MOX" are the analysis results on the mea- [7] and [15], respectively; that of and [13]; that of and [16]; that of + [7]. sured results in the MALIBU program [13,16]. The UO 2 and MOX assemblies were irradiated in Gösgen nuclear power plant in Switzerland.…”

“…Even though C/E-1's obtained by using JENDL-4.0 are not available for the samples in "BWR 9 × 9 UO2 Ac," those for JENDL-3.3 are shown to confirm the trend of the underestimates observed in the burnup analysis with JENDL-3.3. Those shown as "PWR 15 × 15 UO2" and "PWR 15 × 15 MOX" are the analysis results on the mea- [7] and [15], respectively; that of and [13]; that of and [16]; that of + [7]. sured results in the MALIBU program [13,16].…”

“…In recent evaluations of neutron cross sections, the thermal and epithermal cross sections of these isotopes were 2 T. Yamamoto and T. Sakai Even though the isotopic compositions of those nuclides are not large, these differences in the capture cross sections may cause the considerable changes in the isotopic compositions in UO 2 and MOX fuels irradiated in LWRs and in the neutron multiplication factors (k eff 's) of MOX loaded light water cores. The purpose of this paper is to show the effect of the differences in the capture cross sections of 238 Pu on the isotopic compositions of 238 Pu in irradiated UO 2 and MOX fuels irradiated in BWR and PWR cores and that of the differences in the capture cross sections of 238 Pu and 241 Am on the by using the nuclear data libraries published before JENDL-4.0 [7][8][9][10][11][12][13][14][15][16]. The open symbols in Figure 3 show the examples of C/E-1's (calculation/measurement -1) obtained by using JENDL-3.3 as the nuclear data libraries for the continuous energy Monte Carlo burnup calculation code MVP-BURN [17] in the horizontal axis of the measured sample burnups.…”

“…The burnup calculations of MVP-BURN for both libraries were performed by using a general-purpose chain model (u4cm6fp104bp12T for JENDL-3.3 and u4cm6fp110bp14T for JENDL-4.0). The fuel samples are BWR UO 2 fuels [7,15], and PWR UO 2 and MOX fuels [13]. The C/E-1's shown as "BWR 9 × 9 UO2 FP" are on the measured isotopic compositions which were measured by the Japan Atomic Energy Agency (JAEA) under the experimental program which was aimed at the isotopic composition measurements of U, Pu and major fission product nuclides for fuel samples from three-and five-cycle-irradiation BWR 9 × 9 fuel assemblies discharged from Fukushima Daini Unit 1 [15].…”

The Application of Radiochemical Measurements of PWR Spent Fuel for the Validation of Burnup Codes

“…The validation of Japanese SWAT4 burnup code using PWR Mixed Oxide Fuel (MOX) fuel is presented by Kashima [12]. High-burnup PWR MOX and UO 2 fuels were used by Yamamoto for the validation of the MPV-BURN and SRAC burnup code [13]. The importance of burnup calculation reliability for criticality safety and spent fuel disposal is described in the papers by Frankl and Vasiliev [14,15].…”