Katsumi Une

Formation and growth of intragranular fission gas bubbles in UO2 fuels with burnup of 6–83 GWd/t

Abstract The detailed characteristics of intragranular fission gas bubbles in UO2 fuel pellets (burnup: 6–83 GWd/t) before and after postirradiation annealing at 1600 and 1800°C have been examined by TEM and SEM fractography. In the base-irradiated fuels, a high density of small bubbles of about 2 nm in diameter precipitated uniformly in the matrix. When increasing burnup above 44 GWd/t, larger bubbles of 10–20 nm newly appeared in addition to the small bubbles. On heating at high temperatures, bubble growth was saturated within a few minutes. Enhanced coarsening of bubbles was found preferentially near the grain boundaries in the middle burnup fuels of 16–28 GWd/t and throughout the grains in the high burnup fuels of 44 and 83 GWd/t. The bubble growth during annealing was associated with a remarkable decrease of the bubble number density, and the relationship between bubble density Nb and mean diameter db was expressed as Nbαdb−2.6. The coarsening was attributed to coalescence via bubble migration for moderately large bubbles of up to 50–60 nm, and to Ostwald ripening accompanied by a sufficient vacancy supply from external vacancy sources such as free surfaces, grain boundaries or irradiation-induced sub-grain boundaries for huge bubbles above 100 nm.

Microstructural change and its influence on fission gas release in high burnup UO2 fuel

Abstract The microstructural change of UO2 fuel pellets (burnup: 6–83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

Crystallography of zirconium hydrides in recrystallized Zircaloy-2 fuel cladding by electron backscatter diffraction

Precipitation morphology and habit planes of the δ-phase Zr hydrides, which were precipitated within the a-phase matrix grains and along the grain boundaries of recrystallized Zircaloy-2 cladding tube, have been examined by electron backscatter diffraction (EBSD). Radially-oriented hydrides, induced by residual tensile stress, precipitated in the outside region of the cladding, and circumferentially-oriented hydrides in the stress-free middle region of the cladding. The most common crystallographic relationship for both types of the hydrides precipitated at the inter- and intra-granular sites was identical at (0001)α // {111}δ, with {10&1macr;7}α // {111}δ, being the occasional exception only for the inter-granular radial hydrides. When tensile stress was loaded, the intra-granular hydrides tended to preferentially precipitate in the grains with circumferential basal pole textures. The inter-granular hydrides tended to preferentially precipitate on the grain faces opposite to tensile axis. The change of prioritization in the precipitation sites for the hydrides due to tensile stress could be explained in terms of the relaxation effect of constrained elastic energy on the terminal solid solubility of hydrogen at hydride precipitation.

Effects of additives and the oxygen potential on the fission gas diffusion in UO2 fuel

Diffusion coefficients, D, of 133Xe have been measured in the temperature range 1000–1600°C for UO2, 0.5 wt% Nb2O5-UO2 and 0.2 wt% TiO2-UO2 with the irradiation dose of 1.2 × 1017fissions/cm3 (4 MWd/t U), by means of a post-irradiation annealing technique. In addition, the effect of the oxygen potential on the diffusion coefficient for UO2 has been examined at 1400°C over the range of −560 to −390 kJ/mol. The diffusion coefficients are enhanced by the doping of Nb2O5 and TiO2 by a factor of about 50 and 7, respectively. The results for the three specimens at the oxygen potentials corresponding to a constant PH2O/PH2 ratio of 1.1 × 10−3 are as follows: D(cm2/s) = 4.0 × 10−8exp(−264(kJ/mol)/RT) for UO2, D(cm2/s) = 4.6 × 10−5exp(−306(kJ/mol)/RT) for 0.5 wt% Nb2O5-UO2, D(cm2/s) = 5.0 × 10−7exp(−272(kJ/mol)/RT) for 0.2wt% TiO2-UO2. The diffusion coefficient for undoped UO2 is increased by a factor of about 3 when increasing the oxygen potential from −556 to −391 kJ/mol at 1400°C. Over this range of the oxygen potential, the corresponding O/U ratio changes from 2.000 to 2.001. The present tendency is in agreement with previously reported results on cation self-diffusion and diffusion controlled phenomena such as creep and grain growth in undoped and additive doped UO2 fuels.

Oxygen Potentials of UO2 Fuel Simulating High Burnup

The oxygen potentials at 1,000 and 1,300°C and the lattice parameters of UO2 fuels with soluble fission product elements (Zr, Ce, Pr, Nd, Y), simulating high burnup of up to 10a,o have been measured by means of thermogravimetry and X-ray diffraction. The oxygen potentials for (U, FP)O2+x fuels are higher than pure UO2+x; at a given composition and increase positively with increasing simulated burnup. They can be represented as a function of the mean uranium valence at compositions of 0/M>2.01. The lattice parameters of stoichiometric (U, FP)02.00 fuels decrease linearly with simulated burnup, and can be expressed as a (pm) = 547.02–0.1225, where B is burnup in a.o

Effects of Temperature Cycling and Heating Rate on Fission Gas Release of BWR Fuels

The effects of temperature cycling and heating rate on the release behavior of 85Kr have been studied for U02 pellets irradiated in a commercial BWR during 3 and 4 cycles (burn-up: 23 and 28GWd/t), by using a post irradiation annealing technique. In addition, characteristics of intergranular bubbles in base-irradiated and annealed specimens (burn-up: 6~28GWd/t) have been examined by SEM fractography. No significant difference in the release of 85Kr was observed between the cyclic heating from 700 to 1,400°C and isothermal heating at 1,400°C. The maximum release rate of 85Kr during heating up to 1,800°C became lower with decreasing heating rate in the range of 0.03–10°C/s, while its cumulative fractional releases were about 20~30%, almost independent of heating rate. The fractional coverage of the grain face area occupied by intergranular bubbles saturated around 40~50 for the specimens annealed at 1,600-1,800°C, independent of specimen burn-up and heating conditions (temperature, heating rate and duration...

Effect of sintering atmosphere on the densification of UO2Gd2O3 compacts

Abstract Sintering kinetics of powder compacts of UO 2 -(5,10 wt%)Gd 2 O 3 and UO 2 have been studied in controlled atmospheres of H 2 O / H 2 and CO 2 / CO mixed gases by using a dilatometer. The densification rates and microstructure of the sintered pellets are considerably influenced by both the sintering atmosphere and Gd 2 O 3 content. After a heat treatment of 1650° C for 2 h, the sintered densities for UO 2 -Gd 2 O 3 pellets begin to decrease above the threshold oxidizing atmospheres, while the density for the UO 2 pellet increases slightly with more oxidizing atmospheres. These behaviors result from the difference in development of pore structure during sintering: the pore structure of UO 2 -Gd 2 O 3 pellets varies from an open pore structure to a closed pore structure on changing the sintering atmosphere from reducing to oxidizing. On the other hand, the pore structure of the UO 2 pellet is hardly affected by the sintering atmosphere. The formation of (U,Gd)O 2 solid solutions and the grain growth are enhanced with more oxidizing atmospheres.

Terminal Solid Solubility of Hydrogen in Unalloyed Zirconium by Differential Scanning Calorimetry

Zircaloy-2 cladding lined with pure Zr (Zr liner cladding) was developed as a candidate for pellet-clad interaction/ stress-corrosion cracking (PCI/SCC) failure resistant BWR cladding, and its good performance has been proved in many power ramp tests, up to burnups of about 50GWd/t. However, at higher burnups above 50–60GWd/t, another type of PCI failure, which apparently possesses an outsidein type cracking, has been recently reported. Though a hydride precipitation assisted-failure mechanism has been proposed for this type of fuel failure, there is no established theory. When considering the fuel performance, two phenomenological hydride behaviors seem to be important: (1) a dynamic relationship between hydride precipitation at crack tips and crack propagation in Zircaloy, and (2) a heterogeneous accumulation of hydrides in the Zr liner region adjacent to the Zr liner/Zircaloy-2 interface, which was observed in high burnup BWR claddings after base irradiation. The latter phenomenon may influence on hydride re-precipitation in the cladding outside region at power ramp conditions. In order to clarify both phenomena, basic understanding of dissolution and precipitation behavior of hydrides not only in Zircaloy-2 but also in Zr liner is essential. The terminal solid solubility during the dissolution of hydrides (TSSD) at heatup and during the precipitation of hydrides (TSSP) at cooldown have been the subjects of many reports for Zr and its alloys, using various measuring techniques. However, there is considerable scatter in the published data, especially in TSSP data. Moreover, there are no systematic data sets of TSSD and TSSP solvi for Zircaloy-2 and unalloyed Zr using the same technique. In our previous study, we have derived data sets of TSSD and TSSP for the current Zircaloy-2 and improved High Fe Zircaloy for BWRs, by using differential scanning calorimetry (DSC), and best fit equations for the two solvi were presented. In this subsequent study, the same DSC technique was applied to obtain data sets of TSSD and TDDP for unalloyed Zr. The present data for unalloyed Zr were compared to the previous data for the two types of Zircaloys. 1. Experimental (1) Materials In this study, two types of unalloyed Zr were subjected to terminal solid solubility (TSS) measurements. One was current Zr liner material, which was directly prepared from a tube shell of Zr liner cladding with 72.5mm outside and 41.5mm inside diameter. The shape of the Zr liner specimens was a 4mm square, about 0.6–0.7mm thick and weighing about 60–70mg. The other Zr specimens with almost the same shape as the Zr liner specimens were prepared from a Zr slab, supplied by National Bureau of Standards (NBS). Table 1 shows impurity concentrations of the two types of unalloyed Zr specimens, which were used for the present TSS measurements. Gaseous impurities of H, N and O in the Zr liner specimens were somewhat lower than those in the NBS Zr specimens. The oxygen concentrations were 320 and 850 ppm for the former and latter specimens. Test samples were hydrogenated from the as-received hydrogen levels (Zr liner: 9 ppm; NBS Zr: 27 ppm) by (1) a corrosion reaction in water vapor of 9.6MPa at 400 C, or (2) by a gaseous hydrogenation at 500 C in a He/2%H2 mixed gas. The obtained hydrogen concentration ranged from the asreceived levels to 281 ppm. The details of the hydrogenation procedure were described previously. (2) Differential Scanning Calorimetry The TSSD and TSSP temperatures of the specimens were measured using the differential scanning calorimetry (DSC) technique. The details of the DSC instrument (Netzsch DSC-404) were described previously. The DSC measurements were carried out in the same manner as the previous study for Zircaloy-2 and High Fe Zircaloy. Namely, carrier gas was purified Ar at the flow rate of 50 cm/min, and the heatup and cooldown rate of 10 C/min was adopted for the maximum temperature of 500 C with a 5min holding period there. The first run data were excluded, because of inhomogeneity of hydrogen distribution in the specimens. Then the data of the subsequent second and third runs were adopted for evaluating DSC peaks. The analysis of DSC peaks resulting from hydride dissolution during heatup and hydride precipitation during cooldown followed the previous procedure.

Rim structure formation of isothermally irradiated UO2 fuel discs

UO2 fuel discs, which had been irradiated at an isothermal condition of 550-630 degreesC to 51, 86 and 90 GWd/t without any restraint pressure, have been subjected to detailed microstructure observations. elemental analyses and density measurements. Their data were compared with previously reported results of high-burnup Zircaloy-clad type fuel pellets, so as to clarify the effect of pellet-cladding interaction (PCI) restraint on the rim structure formation. The porous rim structure, accompanying huge bubbles and high porosity, was recognized for the high-burnup discs of 86 and 90 GWd/t, but not for the 51 GWd/t disc. From the good coincidence between porosity increase and density decrease for the high-burnup discs, it was concluded that the precipitation and growth of coarsened rim bubbles substantially caused fuel swelling. The wide variety of rim bubble sizes and porosities at a given local burnup, which was recognized in the present PCI-free discs and Zircaloy-clad type pellets reported in other literature. possibly resulted from the external PCI restraint effect. The pressure difference between bubble internal and external pressures, and vacancy diffusivity would rate-control the growth of rim bubbles

Fission gas release during postirradiation annealing of UO2-2 wt% Gd2O3 fuels

Abstract UO 2 -2 wt% Gd 2 O 3 fuels (grain size: 4 μm) irradiated in a commercial BWR during 1, 2, 4 and 5 cycles (burnup: 4–38 GWd/t) have been annealed at 1800°C in an out-of-pile condition and the release rate of 85 Kr was measured continuously, so as to examine the burst release during temperature ramp and diffusional release during isothermal annealing. The data were compared with those of UO 2 fuels (grain size: 9 μm) obtained previously. In addition, the change of fuel microstructure brought about by the annealing was examined. The critical temperature for the onset of burst release of 85 Kr for UO 2 -2 wt% Gd 2 O 3 fuels became lower with increasing burnup, and constant between 1400 and 1500°C at higher burnups above 20 GWd/t, as in the case of UO 2 fuels. The fractional burst releases for UO 2 -2 wt% Gd 2 O 3 were about 2–3 times larger than the values of UO 2 at a given burnup. Effective diffusion coefficients of 85 Kr for UO 2 -2 wt% Gd 2 O 3 fuels were almost equivalent to the values of UO 2 fuels and appeared to be independent of burnup. The bubble swelling caused by annealing at 1800°C for 5 h increased with burnup. and was about 1.5 times larger than that of UO 2 fuels at high burnups. The remarkable differences in fractional burst release and bubble swelling between UO 2 -2 wt% Gd 2 O 3 and UO 2 fuels were mainly due to the difference in grain size of both fuels.