Chemical state of fission products in irradiated UO2

The chemical state of fission products in irradiated UO₂ fuel has been estimated for FBR as well as LWR on the basis of equilibrium calculation with the SOLGASMIX-PV code. The system considered for the calculation is composed of a gas phase, a CaF₂ type oxide phase, three grey phases, a noble metal alloy, a mixed telluride phase and several other phases each consisting of single compound. The distribution of elements into these phases and the amount of chemical species in each phase at different temperatures are obtained as a function of oxygen potential for LWR and FBR. Changes of the chemical potential of the fuel-fission products system during burnup are also evaluated with particular attention to the difference between LWR and FBR. Some informations obtained by the calculation are compared with the results of post irradiation examination of UO₂ fuels.     

Behavior of Fission Products Zirconium and Barium in Fast Reactor Fuel Irradiated to High Burnup

Barium and Zr generated in nuclear fuels can precipitate as multi-component oxide with some other fission products. In addition, the solubility of Ba in the fuel depends on the oxygen potential and the temperature and Zr can easily dissolve into the fuel matrix. Therefore, the behavior of the Ba-Zr oxide inclusions during irradiation is rather complex. In this work, the composition of multi-component oxides and the distributions of Ba and Zr as a function of relative radius were evaluated with X-ray microanalysis. As results, the oxide inclusions containing both Ba and Zr and containing only Ba were observed in the fuel irradiated to the burnup of 13.3 and 10.6 at%, respectively. These results were discussed in terms of the solubility of Ba and Zr in the fuel and in terms of the rO₂–UO₂ phase diagram, together with the radial distributions of Ba and Zr in fuel matrix.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

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.     

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.     

Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

We report the measurement of elastic constants of non-irradiated UO₂, SIMFUEL (simulated spent fuel: UO₂ with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young’s modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.     

Distribution of Fission Products in Irradiated UO2

Inhomogeneities of fission products and plutonium distribution in irradiated UO₂ were examined. UO₂ pellets were ground so as to let them fracture along the grain boundaries, and the fragments were then dissolved successively into several fractions. The specific activities of these fractions were compared, and the distribution of fission products in the grain and grain boundaries were studied. It was found that at lower temperatures of irradiation (～1,400°C), the fission products, except Cs, were distributed fairly uniformly over the grain, while Cs had accumulated at the grain boundaries. At higher temperatures (near or above the melting point of UO₂), inhomogeneity was noticed in the Zr-Nb and Cs distribution. The concentration of these elements varied along the temperature gradient of the fuel rod than within the grains. Zr-Nb was found to be concentrated in the higher temperature areas of the fuel rod, while conversely, the concentration of Cs was smaller in the same areas. Distribution of Ce, Ru and Pu was found to be fairly uniform throughout the grains. The process of fission product migration in UO₂ is also discussed in the paper.     

Effect of Irradiation on the Microstructure of Swaged UO2

A study was made of the microstructure of swaged UO₂, irradiated to a burnup of 3,000 MWD/T-U at an estimated center-line temperature of 1,400°C.

In the peripheral zones of the specimen near the cladding, evidence was observed of sintering of the UO₂ particles by irradiation even at comparatively low temperatures (about 300°–400°C). Tracks of fission product fragments were observed at the inner wall surface of cracks in the specimen. The width of the tracks is seen to be larger than observed on thin films and on sintered UO₂. Mixed phases of UO₂ and fission products precipitated in the interior of crystal grains, observed in the form of hillocks left unetched, were surmounted by several sprouts of substances still more chemically stable. Mixed phases of UO₂ and fission products were precipitated on the inner wall surface of the cracks and appeared upon etching as stepped protuberances of irregular contours. A comparison between the microstructures of swaged UO₂ and sintered UO₂ irradiated under the similar conditions is discussed.     

Re-solution-controlled fission-gas behavior in UO2 irradiated in a fast flux

Uranium dioxide irradiated in a fast neutron flux to a burnup of $\text{2 × 10}{}^{20}$ fissions/cm³ between 650 and 1400°C has been examined by transmission- and scanning-electron microscopy and replication metallography. The fission-gas distribution in the fuel matrix and grain boundaries has been characterized as a function of irradiation temperature and fission rate. The majority of fission gas produced even at the highest irradiation temperature was in the UO₂ matrix either in solution or in the form of bubbles < 20 Å in diameter. The results are explained on the basis of an irradiation-induced re-solution mechanism whereby fission gas from within bubbles is reinjected into metastable solution in the UO₂ lattice. Calculated fission-gas solubilities are given as a function of temperature for 10¹³, $\text{3 × 10}{}^{13}$, and 10¹⁴ fissions/cm³ · sec, and, based on these results, it is concluded that the re-solution process is operative over a substantial fuel volume of both light-water-reactor and fast-breeder-reactor oxide fuels.     

Burnup Analysis of Uranium Dioxide Nondestructive Analysis by External Conversion Method

Fundamental experiments on the nondestructive burnup analysis of UO₂ were made by external conversion method using a sector type double focusing β-ray spectrometer. Details of the procedure are described. About 100 mg of UO₂ enclosed in an Al-tube was irradiated and analyzed by this method. The fission products used as burnup indicators were ⁹⁵Zr, ⁹⁵Nb, ¹⁴⁰La and ¹⁰³Ru. A code for fission product calculation developed in our laboratory was used in burnup determination. The burnup thus determined was 53.07 MWD/t.     

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.     

The radial distribution of fission gases and other fission products in irradiated PWR fuel

The radial distribution of fission gas (xenon) and other fission products (cesium, ruthenium, cerium) has been measured on UO₂ fuel pellets irradiated in commercial pressurized water reactors to burnups between 13.23 and 48.26 GWd/tU. Fission gas release occurs from the pellet center, and at temperatures $\text{< 1300° C}$ is confined to the region of grain growth. The maximum fractional release measured at the center ranges from 20% to 30%. Only at high burnup (48.26 GWd/tU) an additional release of cesium has been observed. This is considered as evidence for an increase in fission product release at higher burnups. At fuel center line temperature $\text{> 1500° C}$ a high fission gas release is accompanied by a high cesium release. The local release starts at the onset of fission gas bubbles precipitating on grain boundaries and saturates in the center of the pellet at a fractional release value of about 90%.     

Formation of Pellet-Cladding Bonding Layer in High Burnup BWR Fuels

Formation process of the pellet-cladding bonding layer was studied by EPMA, XRD, and SEM/TEM for the oxide layer on a cladding inner surface and the bonding layer in irradiated fuel rods. Specimens were prepared from fuels which had been irradiated to the pellet average burnups of 15, 27 and 42 GWd/t in BWRs. In the lower burnup specimens of 15 and 27GWd/t, no bonding layer was found, while the higher burnup specimens of 42 and previously reported 49 GWd/t had a typical bonding layer. A bonding layer which consisted of two regions was found in the latter fuels. One region of the inner surface of the cladding was made up mainly of ZrO₂. The structure of this ZrO₂ consisted of cubic phase, while no monoclinic crystals were found. The other region, near the pellet surface, had both a cubic solid solution of (U, Zr)O₂ and amorphous phase. Even in the lower burnup specimens having no bonding layer, cubic ZrO₂ phase was identified in the cladding inner oxide layer. The formation process of the bonding layer were discussed in connection with phase transformation by irradiation damage of fission products and conditions for contact of pellet and cladding.     

Thermodynamic analysis of uranium–gadolinium fuel stability at high temperatures

Using the most accurate measurements of the liquidus temperature in the UO₂–Gd₂O₃ system up to 30 mol.% of Gd₂O₃, thermodynamic models of the melt and cubic solution GdO_1.5 in UO₂ are constructed. The equilibrium phase diagram of the system UO₂–GdO_1.5 in the interval 1900–3200 K is calculated in the entire composition range and the metastable diagram is calculated assuming that no cubic solid solutions are formed. The upper and lower boundaries of the melting onset temperature (solidus) of uraniumgadolinium fuel are presented. The phase composition of the pellets made from such fuel and, ultimately, the technology determine the melting onset temperature uniquely.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Effect of burn-up on the thermal conductivity of uranium dioxide up to 100.000 MWd t

The thermal diffusivity and specific heat of reactor-irradiated UO₂ fuel have been measured. Starting from end-of-life conditions at various burn-ups, measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage were determined. In this context, particular emphasis was given to the behaviour of samples displaying the high burn-up rim structure. Recovery stages could be thoroughly investigated in samples that were irradiated at low burn-ups and/or at high irradiation temperatures. Other samples, in particular those exhibiting the characteristic rim structure, disintegrated at temperatures slightly higher than the irradiation temperature. Finally, from a database of several thousand measurements, an accurate formula for the in-pile thermal conductivity of UO₂ up to 100 GWd t⁻¹ was developed, taking into account all the relevant effects and structural changes induced by reactor burn-up.     

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 10^a,_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 UO_2+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