P.G. Lucuta

A pragmatic approach to modelling thermal conductivity of irradiated UO2 fuel: Review and recommendations

The thermal conductivity of irradiated UO2 fuel is discussed considering the effects of burnup (dissolved and precipitated solid fission products), porosity and fission-gas bubbles, deviation from stoichiometry and radiation damage based on single-effect results previously published on SIMFUEL (simulated extended burnup UO2 fuel) and on radiation damage measurements. An analytical expression including factors describing the above effects is applied to the expression for unirradiated UO2 thermal conductivity; it reflects the knowledge available today, and it is recommended for use with irradiated fuel. The expression is validated against available published data on thermal conductivity of irradiated fuel. This expression can be incorporated into fuel modelling codes to improve calculations of operating temperatures and predictions of behaviour of irradiated fuel under normal and accident conditions, including the extended burnup.

MICROSTRUCTURAL FEATURES OF SIMFUEL : SIMULATED HIGH-BURNUP UO2-BASED NUCLEAR FUEL

Abstract Simulated high-burnup nuclear fuel (SIMFUEL) replicates the chemical state and microstructure of irradiated fuel so that detailed experiments on fission-gas release, thermal conductivity and leaching can be undertaken in the laboratory. Eleven stable elements were added to simulate the compositions of 3 and 6 at% burnup UO2-based fuel. The preparation route featured high-energy grinding and spray drying to achieve homogeneous dispersion of the feed materials on a submicrometre scale. Sintering at 1650°C then provided atom-scale mixing and second-phase development characteristics of high temperature fuels. The gases and volatiles cannot be introduced during fabrication, but can be subsequently added by ion implantation. The microstructure of SIMFUEL was found to be similar to that of irradiated fuel (without bubbles). Spherical metallic Mo-Ru-Pd-Rh precipitates were found uniformly dispersed throughout the matrix. A finely-precipitated perovskite phase was observed decorating the matrix grain boundaries.

Swift heavy ion and fission damage effects in UO2

Abstract Irradiations of uranium dioxide, UO2 with different swift heavy ions were performed using wide ranges of energies and fluences from 72 MeV to 2.7 GeV in the range 5×109–1017 ions/cm2. The ions were Zn, Mo, Cd, Sn, Xe, I, Pb, Au and U. The threshold energy loss for formation of visible tracks in UO2 could be determined to be in the range 22–29 keV/nm. Fission products of fission energy are below this threshold but nevertheless form thermal spikes in UO2. Observable tracks are only found at the surface. By using 127I-beams of 72 MeV energy the consequences of fission product impact, i.e., lattice parameter increase, fission gas bubble formation, resolution of fission gas from bubbles and fission-enhanced diffusion were all observed and measured. The swelling of UO2 was confirmed to be small and the technologically important process of polygonization (grain subdivision, also called rim-effect in operating UO2-fuel) could be simulated. The results obtained are important in understanding the operating behavior of UO2, today’s fuel in nuclear electricity-producing power stations.

Thermal conductivity of hyperstoichiometric SIMFUEL

Abstract The thermal conductivity of hyperstoichiometric SIMFUEL and UO 2+ x was obtained from thermal diffusivity, specific heat and density measurements. The thermal conductivity of UO 2+ x (no fission products present) decreases with increasing O/U ratio; a reduction of 15%, 37% and 56% at 600°C, and 11%, 23% and 33% at 1500°C, was found for O/U ratios of 2.007, 2.035 and 2.084, respectively. For 3 at% SIMFUEL there was little difference in the thermal conductivity of specimens annealed at oxygen potentials ( ΔG O 2 of -540 kJ/mol (corresponding to UO 2.000 ) and -245 kJ/mol (corresponding to UO 2.007 ). For SIMFUEL, annealed in reducing conditions, the fission products lowered the thermal conductivity significantly. However, for high oxygen potentials ( ΔG O 2 ≥ −205 kJ/mol), the thermal conductivities of UO 2+ x and SIMFUEL were found to be approximately equal in the temperature range of 600 to 1500°C. Consequently, excess oxygen is the dominant factor contributing to thermal conductivity degradation at high oxygen potentials.

Thermal conductivity of SIMFUEL

Abstract Thermal diffusivity and specific heat of SIMFUEL — simulated high-burnup UO 2 fuel with an equivalent burnup of 3 and 8 at% — were measured between 300 and 1800 K, and the data were combined to obtain the thermal conductivities. The thermal conductivity of SIMFUEL provides a model for the intrinsic conductivity of high-burnup fuels (i.e. without gas bubbles). The results on conductivity of 3 and 8 at% burnup SIMFUEL were lower than those measured for UO 2 by 29 and 45% at 300 K and by 6 and 15% at 1773 K. The reduction in thermal conductivity was approximately linear with burnup. The change in thermal conductivity is a possible explanation for the enhanced gas release observed from high-burnup fuel.

Modelling of UO2-based SIMFUEL thermal conductivity The effect of the burnup

Abstract Thermal conductivities of stoichiometric SIMFUEL — simulated high-burnup UO 2 fuel — with equivalent burnups of 1.5, 3 and 8 at%, deduced from thermal diffusivity and specific heat measurements, are modelled as a function of burnup, taking into consideration two key microstructural features of irradiated high-burnup fuel: 1. (i) the precipitated fission-product phases, and 2. (ii) the dissolved fission products in the matrix. The degradation of the UO 2 matrix thermal conductivity due to the dissolved fission products (after corrections for the precipitated-product phases) was analyzed by considering their effect on the phonon heat current. The degree of the scattering of the phonons by dissolved additives in the matrix (scattering parameter) is a function of the phonon frequency, and was determined using phonon heat current theory. The scattering parameter followed the theoretically predicted square root dependence on the temperature and concentration of the dissolved additives. The thermal conductivity of the SIMFUEL matrix, predicted from the model based on the phonon scattering by dissolved fission products taken into account their mass difference, was in good agreement with the measured values. These results indicate that the degradation of fuel thermal conductivity, in the absence of fission-gas bubbles, can be explained by the phonon scattering from the dissolved fission products.

Specific heat measurements of UO2 and SIMFUEL

Abstract New specific measurements are reported for UO 2 with eleven additives, simulating 8 at.% burnup, at temperatures between 300 and 1673 K. A Netzsch DSC 404 heat-flux differential scanning calorimeter was used. The results confirm our earlier measurements, showing no anomalous increase relative to UO 2 .

Radiation damage and simulated fission product effects on the properties of inert matrix materials

Candidate inert matrix materials for actinide transmutation (MgAl 2 O 4 , CeO 2 ) or immobilization (ZrSiO 4 ) containing 241 Am were characterized. The currently most considered material, ZrO 2 , was produced, with La 2 O 3 as stand-in for Am, and with and without simulated fission products to investigate burnup effects. The oxygen potential was measured using an EMF cell. The accumulation of radiation damage due to Am decay was investigated by periodically measuring lattice parameters and hardness. Sequential leaching tests in deionized water, aimed at correlating the leaching behaviour of Am and of the matrix with radiation damage, showed significant release of Am and of some matrix components.