Thermal Conductivity Measurements on UO2+x from 300 to 1,400 K

In order to evaluate the thermal conductivity of oxidized fuel pellets of leaker fuel rods, UO_2+x samples with x between 0.00 and 0.20 were prepared by an oxidation or a sintering method. Sample thermal diffusivities were measured by using a laser flash method from 300 to 1,400 K and their thermal conductivities were evaluated from multiplying the thermal diffusivities by the sample densities and the specific heat capacities derived from the literature. The thermal conductivities of UO_2+x were decreased with increasing hyperstoichiometry and they were expressed as a function of their hyperstoichiometry using the concentration of U⁵⁺ formed with the excess interstitial oxygen atoms.     

Thermal Expansion and Thermal Conductivity of Cesium Uranates

Two kinds of cesium uranates, which are often predicted by thermochemical calculations to be formed in irradiated oxide fuels with high oxygen potentials, were prepared from U₃O₈ and Cs₂CO₃ to determine the thermal expansions and the thermal conductivities. The lattice parameters of tetragonal Cs₂UO₄ and monoclinic Cs₂U₂O₇ were measured by the high-temperature X-ray diffraction method as a function of temperature. The linear thermal expansions of Cs₂UO₄ and Cs₂U₂O₇ obtained from the temperature dependencies of the lattice parameters were 1.2% and 1.1% from room temperature to 973 and 1,073K, respectively. The thermal diffusivities of Cs₂UO₄ and Cs₂U₂O₇ were measured on the disk-shaped samples by the laser flash method as a function of temperature. The thermal conductivities were evaluated from the measured thermal diffusivities and the bulk densities, and the specific heat capacity available in literature. The thermal conductivity of Cs₂UO₄ corrected for 100%TD was 1.2W/m·K at 980K and that of Cs₂U₂O₇ was 0.94W/m·K at 1,093K, which are about 30% and 27% of that of UO₂, respectively.     

Thermal Diffusivities and Thermal Conductivities of UO2-Gd2O3

Thermal diffusivities of samples of UO₂ and UO₂ doped with 3, 5, 7 and 10 w/0 Gd₂O₃ were measured over the temperature range of 298~2,023 K by a laser flash method. Then thermal conductivities were calculated from them. The thermal conductivity decreased with increasing Gd₂O₃ content at low temperatures, while it was independent of Gd₂O₃ content at high temperatures. An expression of the thermal conductivity was proposed for (U, Gd)O₂ solid solution as a function of Gd₂O₃ content and temperature by applying Klemens' model.     

Thermal Conductivities of Granular UO2 Compacts with/without Uranium Particles

The thermal conductivities of granular UO₂ compacts with and without uranium particles were measured to evaluate the thermal performance of vibro-packed granular MOX fuels containing metallic fine particle oxygen getters. The thermal conductivity of the compact with 10 wt% of uranium particles was higher than that of the compact without uranium particles. After heating beyond 1,408 K, the melting point of the uranium particles, the thermal conductivity increased further. The evaluation model for analyzing such phenomena was developed. The model predicted that once the UO₂ compact with uranium particles was exposed to a temperature beyond 1,408 K, the uranium particles should melt and provide interconnecting areas between the UO₂ granules and uranium particles, and between other uranium particles. The resulting increase of the thermal conductivity was reasonably expressed by the effect of necks in the compact on the heat conduction.     

XMA Analysis of Undissolved Materials Observed in NSRR Short Irradiated Uranium Dioxide Fuel

After a short irradiation of UO₂ fuels with and without SiO₂+ Al₂O₃ additive in NSRR, a comparative study of the fission products (FP) dissolved in nitric acid and undissolved materials was performed. Results of the study were: (1) The additive fuel behaved similarly to non-additive fuel. (2) In both types of fuels, the undissolved materials in dried out solution were not FP but metals existing from the fuel fabrication stage. (3) The NSRR short irradiation had great influence on the shape of the undissolved components without changing the as-fabri-cated compositions.     

Melting Temperature of Irradiated UO2 and UO2-2wt%Gd2O3 Fuel Pellets up to Burnup of about 30 GWd/tU

Melting temperature of UO₂ and UO₂-2w/oGd₂O₃ fuels irradiated in a commercial LWR were determined by a thermal arrest technique in a burnup range up to approximately 30GWd/tU.

No decrease in the melting temperature was observed on both UO₂ and UO₂-2w/oGd₂O₃ fuels with increment of burnup to 30GWd/tU. It was also found that the Gd₂O₃ addition below 2w/o has no influence on the melting temperature.     

Effect of Fuel Pellet Eccentricity with Cladding on Fuel Rod Thermal Behavior under Reactivity Initiated Accident Condition

Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g?UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum temperature difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g?UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA conditions.     

Evaluation Methods of Resonance Absorption for System with Pellet Surrounded by Fuel Solution

Abstract

The reaction cross sections of ²⁷Al(n, p)²⁷Mg, ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn, ⁹⁰Zr(n, 2n)^89m+gZr and ⁹³Nb(n, 2n)^92mNb have been measured by the activation method in an energy range of 13.3–14.9 MeV using the intense D-T neutron source, FNS. Absolute flux was determined by the associated α-particle counting method incorporated with neutron spectra obtained from both a Monte Carlo calculation and a time-of-flight measurement. Corrections were extensively performed not only for the neutron flux determination, but also for the low energy neutron contribution to the reaction rates. The present data were compared with comprehensive evaluations as well as recent experimental data. The measured cross sections of ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn and ⁹⁰Zr(n, 2n)^89m+gZr are generally in good agreement within experimental errors with the values in both the JENDL Dosimetry File and IRDF-90. It is also shown that there are the overestimation of the cross sections for ⁹³Nb(n, 2n)^92mNb in the JENDL Dosimetry File, and the over- estimation and underestimation of the cross section for ²⁷Al(n, p)²⁷Mg in the JENDL Dosimetry File and IRDF-90, respectively.     

Dimensional Changes in Irradiated (Th,U)O2

Effects of irradiation on the dimension and microstructure in (Th,U)O₂ pellets were examined by measurements of lattice parameter and bulk density changes, and observations of pore structures. The concentrations of fission-induced defects and the damage volume were estimated by a simple model. Both macroscopic and microscopic dimensional changes were found to increase initially with fission dose and then fall off. The difference between macroscopic and microscopic ingrowths increased with dose, suggesting that fission-induced interstitials would cluster or go to sinks and the concentration of vacancies would be in excess of that of interstititials. The damage volume for vacancies was estimated to be about 1x10^?22m³·fiss.^?1, and almost agreed with that for fission Xe release. Observations of the pore structure indicated that the volume fraction of pores smaller than 2–3 μm decreases with irradiation and the distribution of pore size shifts toward the larger side.     

Effect of Grain Size on Microstructural Change and Damage Recovery in UO2 Fuels Irradiated to 23 GWd/t

The effect of grain size on microstructural change and damage recovery in U0₂ fuels was studied by X-ray diffractometry (XRD) and transmission electron microscopy (TEM). The as-irradiated lattice parameter of the standard fuel (grain size: 16/μm) was larger than that of the large-grained fuel (43 μm), indicating a larger number of fission-induced point defects in the lattice of the former fuel. This tendency was in contrast to previously reported results for low burnup fuels below 1 GWd/t. The lattice dilation in the present high burnup fuels was mainly due to the accumulation of vacancies. The lattice parameter of both fuels began to recover around an irradiation temperature of 450~650°C, and both had a complete recovery at 850°C. On annealing at high temperatures of 1,450~ 1,800°C, the bubble diameter in the standard fuel was larger than that in the large-grained fuel. This indicated that vacancy diffusion from the grain boundaries plays an important role during bubble coarsening at high temperatures.