Effect of Pu-rich agglomerates in MOX fuel on reactivity analysis of light water reactor MOX core physics experiments

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) on the reactivity analysis of light water reactor MOX core physics experiments was studied with the continuous-energy Monte Carlo calculation code MVP II. First, the following three different models were compared in the analysis of a representative unit cell of a MOX core tested at the KRITZ reactor: a Lattice model where Pu-rich agglomerates were assumed to exist in a fixed pitch, a statistical geometry (STG) model of MVP II, and a Random model where the random distribution of Pu-rich agglomerates was directly modeled. Since the three models gave comparable results, the STG model was used in parametric calculations to systematically understand the reactivity effect depending on the characteristics of Pu-rich agglomerates. In addition, the selected unit cells composing the MOX cores and one representing MOX core tested at the EOLE criticality facility were analyzed with the measured characteristics of Pu-rich agglomerates in MOX fuel. Consequently, the reactivity differences between the calculations assuming the homogeneous Pu distributions and those considering Pu-rich agglomerates were less than 0.0005 Δk/k/k', indicating that the effect of Pu-rich agglomerates was small on the reactivity analysis of the MOX cores tested in the EOLE facility.     

Analysis of VIP-BWR reactor core physics experiments on UO2 and MOX mockup fuel assemblies - part 2: analysis with SIMULATE5

Analysis of the three test cores in the VIP-BWR program was performed in a two-dimensional geometrical model with CASMO5 coupled with the JENDL-4.0-based neutron data library, and reported in the previous paper. Following the study, interpretation of the experiments were carried out in a three-dimensional geometrical model with SIMULATE5 for the code validation study. The nuclear libraries for the SIMULATE5 calculations were generated with CASMO5 with the JENDL-4.0-based neutron data library. The effective multiplication factors of the critical cores ranged from 0.9983 to 1.0023 with measurement uncertainties of 0.0003 to 0.0004 (one σ). The root mean squares of (the calculated/the measured-1) for the fission rates at the core-mid plain of all the measured fuel rods were about 3% for the three cores. It was noticed that the calculations underestimated the fission rates of the UO₂ fuel rods and overestimated those of the MOX fuel rods for the test cores loaded with MOX fuel rods, which was consistent with trends in the preceding analysis studies of the VIP-BWR program and other MOX core experiments, and the biases were confirmed in the calculation results of power distributions in MOX-fueled light water reactor cores.     

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (k_eff's) to neutron cross sections used for the reactivity analysis of UO₂ and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in k_eff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO₂ core is mainly attributed to the decreases in the capture cross sections of ²³⁸U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in k_eff are decreases in the capture cross sections of ²³⁸U,²³⁸Pu, ²³⁹Pu, and those to decrease in k_eff are decreases in ν of ²³⁹Pu and increases in the capture cross sections of²⁴¹Am. They compensate each other, and the difference in k_eff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.     

Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors

Measured isotopic compositions of UO₂ and MOX fuel samples taken from irradiated light water reactor fuel assemblies were analyzed by CASMO5 coupled with a JENDL-4.0 base library to assess the uncertainties in the calculated isotopic compositions on heavy and fission product nuclides. The burnup calculations for the analysis were performed based on a single-assembly model taking into account the detail fuel assembly specifications and irradiation histories. For the MOX fuel samples, a multiple-assembly model was also adopted taking into account the effect of the surrounding UO₂ fuel assemblies. The average and standard deviation of the biases (C/E ? 1's (here C and E are calculated and measured results, respectively)) were calculated for each nuclide separately on the PWR and BWR UO₂ fuel samples. The averaged biases for ²³⁵U, ²³⁶U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu were 2.7%, ?0.9%, 0.3%, 0.7%, ?2.4% and ?1.7% for PWR UO₂ samples, and 6.7%, ?1.5%, 2.5%, ?0.6%, 0.4% and ?0.1% for BWR UO₂ samples, respectively. The biases with the single-assembly model on the MOX fuel samples showed large positive values of ²³⁹Pu, and application of the multiple-assembly model reduced the biases as reported in our previous studies.     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.