Masaki Amaya

Heat capacity measurements on unirradiated and irradiated fuel pellets

Abstract Heat capacities (Cp) of oxide fuels are essential for the evaluation of fuel temperature at normal, transient and accidental conditions of light water reactors. In this study, the effect of soluble fission products (FPs) was first examined, and then the Cp measurement technique was improved to obtain reliable data for irradiated pellets of smaller than the standard size. A differential scanning calorimeter was applied to the Cp measurements. From the measurements of standard size α-Al2O3, the inaccuracy of the apparatus was estimated to be about 4%. The Cp data of undoped and simulated FPs-doped UO2 pellets of standard size were measured in the temperature region from 325 to 1673 K. From these results, the effect of impurities on the Cp values of UO2 pellets and the presence of the Cp anomalies in the simulated FPs-doped UO2 pellets were investigated. For measurements of small size specimens, the sample-side crucible was modified. The inaccuracy of the apparatus was estimated to be 5% using a small size of the α-Al2O3 sample. This improved technique was applied to obtain the Cp data of UO2 and UO2–10 wt%Gd2O3 specimens with small sizes from 325 to 1673 K, which had been irradiated at isothermal conditions in a test reactor.

Fuel oxidation and irradiation behaviors of defective BWR fuel rods

The fuel oxidation of UO 2 pellets in two types of defective Zircaloy-clad BWR fuel rods with small leaks has been examined along both pellet diametral and fuel rod axial directions. The post-defect irradiation time was a few months for the base-irradiated full length rod and several minutes for the power-ramped segment rod. No phase change to higher order oxides of U 4 O 9 or U 3 O 8 was found, but hyperstoichiometric UO 2+x with fluorite structure was still present for both fuels. The fuel oxidation significantly depended on the defect size and distance from the defect. The pellet volume-averaged O/M ratios at various axial locations were in the range of 2.02-2.06 for the base-irradiated fuel, and about 2.01 for the power-ramped fuel. The data revealed that pellet oxidation by steam proceeded notably even in a short period of several minutes and played a more important role for generating liberated hydrogen, which could cause secondary hydriding of Zircaloy cladding, in comparison with the inner wall oxidation of cladding

Thermal conductivity measurements on 10 wt% Gd203 doped U02 + x

Abstract In order to evaluate the thermal conductivity of oxidized fuel pellets of leaker fuel rods, 10 wt% Gd 2 O 3 doped UO 2 + x samples with x between 0 and 0.15 were prepared by the oxidation method and their thermal conductivities were estimated from their thermal diffusivity measurements. The thermal conductivity decreased with increasing hyperstoichiometry. The summation rule was valid for the phonon scattering parameter of U 5+ ions and that of Gd 3+ ions. The thermal conductivities of (U, Gd)O 2 + x pellets were evaluated as a function of their hyperstoichiometry.

Heat capacity measurements of U1-yGdyO2 (y=0-0.27) from 325 to 1,673 K

Molar heat capacities were measured for U1-yGdyO2 (y=0–0.27) by differential scanning calorimetry (DSC) in the temperature region from 325 to 1,673 K. The molar heat capacities of U1-yGdyO2 (0<y&;lt;0.27) were close to, or slightly lower than, the values calculated by using the summation rule, and anomalous increases of heat capacities were not clearly observed at temperatures above 1,000 K. The heat capacity analysis for (U, Gd)O2 showed that the excess heat capacities of (U, Gd)O2 tended to decrease with increasing cation concentration excluding U4+ ions in crystals. This suggested that the Schottky term contribution to the heat capacity, which was mainly caused by the excitation of f electrons in the 5f 2 configuration in U4+ ions, decreased in (U, Gd)O2, compared to that in undoped UO2 due to the formation of U5+ and Gd3+ ions in (U, Gd)O2 by addition of Gd2O3. Thermal functions of (U, Gd)O2 were also evaluated.

The effects of oxidation on the thermal conductivity of (U, M)O2 pellets (M = Gd and/or simulated soluble FPs)

Abstract 10 wt% Gd 2 O 3 doped UO 2 + x samples with x being between 0 and 0.15 and simulated soluble FP doped UO 2 + x (simulated burnups: 30 and 60 GWd/tU), with x between 0 and 0.02, were prepared with an oxidation method. Sample thermal diffusivities were measured by using a laser flash method from 300 to 1400 K and sample thermal conductivities were evaluated by multiplying the thermal diffusivities by the sample densities and the specific heat capacities derived from the literature. The thermal conductivities of (U,M))O 2 + x (M = Gd and/or simulated soluble FPs) decreased with increasing hyperstoichiometry and they were expressed as a function of hyperstoichiometry and the concentration of impurities such as Gd 3+ and FP ions.

SiC coating as hydrogen permeation reduction and oxidation resistance for nuclear fuel cladding

Silicon carbide (SiC) coating is one of the countermeasures for the prevention of oxidation and hydrogen embrittlement of fuel claddings because SiC has high resistance of oxidation and hydrogen permeation. Hydrogen permeation and oxidation experiments for the cladding materials with SiC coatings were conducted in unirradiated conditions. The sputtering method was employed to make SiC coatings. In the hydrogen permeation experiment, 316 type of stainless steel (SS316) was used as a base material of the coating. SS316 with SiC coatings showed hydrogen permeation reduction by one order of magnitude. In the oxidation experiments, Zircaloy 4 (Zry-4) and SS316 were used as base materials of the coatings. The weight gain of the Zry-4 specimens with a SiC coating decreased by about one-fifth compared to the uncoated ones at 750 °C and 1200°C. This phenomenon was observed for SS316 at 750 °C as well. The peel-off of the coating was observed in several experiments, and it is considered that the peel-off was caused by the difference of the thermal expansions between coatings and base materials. Thicker coatings showed better oxidation resistance, but thinner coatings showed more tolerance of peel-off.

Thermal Conductivity Change in High Burnup MOX Fuel Pellet

High burnup MOX and UO2 test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO2 test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO2 fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.

Measurements of Crystal Lattice Strain and Crystallite Size in Irradiated UO2 Pellet by X-ray Diffractometry

Lattice parameters, crystallite sizes, and nonuniform strains of high-burnup UO2 fuel samples were measured using micro-X-ray diffractometry in order to investigate the effects of grain subdivision and strain distribution between crystallites on the microstructural changes, so-called rim structure formation, in UO2 pellets. The samples were prepared from fuel rods irradiated in commercial reactors, and pelletaveraged burnups of the samples were in the range from 41 to 61GWd/t. While the lattice parameters of the samples increased linearly in the burnup region up to approximately 70 GWd/t, the lattice parameters slightly decreased and tended to level off above 70GWd/t. The measured crystallite sizes were in the range from 100 to 200 nm and these were nearly the same as those of the “recrystalized grains” in the rim structure. The elastic strain energy densities, which were evaluated from the lattice parameters and nonuniform strains, tended to increase with burnup and show two plateaus at burnups of approximately 50 and 70GWd/t, respectively. The saturation of the measured strain energy density at approximately 50GWd/t can be attributed to the tangle of accumulated dislocations, and the increase in the strain energy density above 60GWd/t can be explained by the strain caused by crystallite rotation under the restraint conditions between crystallites.

The effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

In order to investigate the effect of oxidation and crystal phase condition on the ballooning and rupture behaviors of cladding tube under simulated loss-of-coolant-accident conditions, laboratory-scale experiments were performed in which internally pressurized non-irradiated Zircaloy-4 (Zry-4) cladding specimens were heated to burst in steam and argon gas conditions. Values of the maximum circumferential strain were normalized by dividing them by engineering hoop stress at the time of rupture. The dependence of the normalized value on burst temperature and the relationship between the normalized value and the length, width and area of rupture opening were evaluated. The correlation between the normalized value and the burst temperature suggested that the fraction of the β-phase in Zry-4 cladding specimens affected the strain in the specimens and the oxidation of specimens suppressed the amount of ballooning of the specimens. The relationship between the normalized value and the length, width and area of rupture opening indicated that the length, width and area of rupture opening depended on the crystal phase condition in Zry-4 cladding specimens irrespective of atmosphere in the case of the heating rate of ∼3 K/s.

Improved-EDC tests on the Zircaloy-4 cladding tube with an outer surface pre-crack

ABSTRACT In order to investigate the failure behavior of fuel cladding under a reactivity-initiated accident (RIA) condition, biaxial stress tests on unirradiated Zircaloy-4 cladding tube with an outer surface pre-crack were carried out under room temperature conditions by using an improved Expansion-Due-to-Compression (improved-EDC) test method which was developed by Japan Atomic Energy Agency. The specimens with an outer surface pre-crack were prepared by using Rolling-After-Grooving (RAG) method. In each test, a constant longitudinal tensile load of 0, 5.0 or 10.0 kN was applied along the axial direction of specimen, respectively. All specimens failed during the tests, and the morphology at the failure opening of the specimens was similar to that observed in the result of post-irradiation examinations of high burnup fuel which failed during a pulse irradiation experiment. The longitudinal strain (ϵtz) at failure clearly increased with increasing longitudinal tensile loads and the circumferential strain (ϵtϑ) at failure significantly decreased in the case of 5.0 and 10.0 kN tests, compared with the case of 0 kN tests. From these tests, the data of cladding failure were obtained in the range of strain ratio (ϵtz/ϵtϑ) between about −0.6 and 0.7: this range of strain ratio covers the range between about 0.0 and 0.7 which is estimated in the case of RIA-simulated test. It is considered that the data obtained in this study can be used as a fundamental basis for quantifying the failure criteria of fuel cladding under a biaxial stress state.