C.T. Walker

The radial distribution of plutonium in high burnup UO2 fuels

Abstract A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21000 and 64000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions.

Modelling the high burnup UO2 structure in LWR fuel

The concept of a burnup threshold for the formation of the high burnup UO2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60–75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U.

EPMA and SEM of fuel samples from PWR rods with an average burn-up of around 100 MWd/kgHM

Samples from the high and low power regions of two fuel rods with average burn-ups of 89.5 and 97.8 MWd/kgHM were examined. Electron probe microanalysis (EPMA) was used to measure the radial distributions of Xe, Cs and Nd in the UO2 fuel matrix, and scanning electron microscopy (SEM) was used to study the change in the UO2 microstructure across the fuel pellet radius. EPMA showed that a large fraction of the xenon created had been released from the fuel matrix of the high power samples at all radial positions. In these samples, SEM revealed the presence of recrystallised grains at intermediate radial positions where thermal fission gas release had previously occurred. Evidence of recrystallisation was found throughout the pellet cross-sections of the low power samples. It is concluded that recrystallisation of the fuel grains at intermediate radial positions is mainly responsible for the marked increase in fission gas release to the rod free volume at burn-ups above 80 MWd/kgHM.

Oxide fuel transients

Abstract The behavior of UO 2 fuel from power reactors has been studied up to high burn-up (~ 60 GWd/tU), under both steady state and power transient conditions in the range of 35 to 48 kW/m. Careful post-irradiation examination involving electron probe microanalysis, transmission and scanning electron microscopy in addition to detailed hot cell examination have provided a large data base on the radial migration and release of volatile fission products. Theoretical evaluation with different computer codes and supporting laboratory experiments established the basis for understanding the kinetics and mechanisms operative during transients in high burn-up UO 2 . Typical examples are given to demonstrate the degree of understanding achieved

Concerning the microstructure changes that occur at the surface of UO2 pellets on irradiation to high burnup

Abstract It is shown that in addition to the precipitation of small gas filled pores, there is a pronounced reduction in the grain size at the surface of UO2 fuel at high burnup. These microstructure changes were first observed when the local burnup exceeded 70–80 MWd/kgM. Generally, the change in microstructure does not penetrate more than 200 μm. However, in a HWR fuel irradiated to 75 MWd/kgM a large part of the pellet cross section was found to have been affected. Temperature predictions for this fuel suggests that it is the restructuring accompanying thermally activated fission gas release at 1100 to 1200°C that limited the distance over which the microstructure changes occur. Apparently, the formation and fission of Pu is not directly responsible for the change in fuel microstructure. The porosity evidently contains part of the fission gas that is lost from the UO2 lattice in the region where the microstructure changes take place.

Transmutation of neptunium and americium in a fast neutron flux: EPMA results and KORIGEN predictions for the superfact fuels

In the Superfact Experiment four oxide targets containing high and low concentrations of 237Np and 241Am and representing the homogeneous and heterogeneous in-pile recycling concepts were irradiated in the PHENIX reactor. The burnup reached 6.4% FIMA in the targets with low concentrations of Np and Am and 4.5% FIMA in the targets with high concentrations of Np and Am. About 25% of initial concentration of 237Np and 241Am was transmuted. Generally, the radial distribution of Np and Am was quite flat indicating an even rate of transmutation over the pellet cross section. In the targets with 45% and 20% Np, 10 and 12 wt% Pu was created; most of this was 238Pu with a half-life of 88 y. All the targets exhibited high fission gas release of 67 to 77%. As with standard LMFBR oxide fuel, Cr2O3 was the main product of fuel-cladding chemical interaction. In the target containing 20% Am, an FePd alloy was present in all the major radial cracks.

Concerning the development of grain face bubbles and fission gas release in UO2 fuel

Abstract A series of scanning electron micrographs trace the development of gas bubbles on the grain faces in a transient tested BWR fuel. The observations indicate that gas bubbles nucleate on metallic precipitate particles and that bubble migration is prevented by the precipitates. X-ray fluorescence and electron probe microanalysis results show an accumulation of Xe on the grain boundaries at intermediate radial positions and high “in grain” retention in the centre of the fuel. These findings are ascribed to high compressive forces in the fuel during the transient and an increase in the rate of thermal resolution with temperature. It is concluded that under transient conditions the magnitude of the local mechanical restraint and the existence of macroscopic pressure gradients are important in determining the behaviour of grain face bubbles. In a high pressure field both the bubbles and precipitates grow by Ostwald ripening and little bubble interlinkage occurs. In a non-uniform pressure field bubble interlinkage can take place in areas where the restraint pressure is low and this leads to the formation of escape tunnels at the grain edges. Intergranular cracks which form at the end of the transient when the mechanical restraint pressure is removed also provide escape routes for fission gas.

Temperature measurements in high burnup UO2 nuclear fuel: Implications for thermal conductivity, grain growth and gas release

Abstract Fuel centreline temperatures measured under transient conditions are reported. These show that also at burnups as high as 4.5% FIMA the onset temperature for thermal gas release is close to 1200°C. At constant linear power the 1200°C isotherm moves outwards with burnup due to the development of a thermal barrier at the pellet rim and a general degradation in the fuel thermal conductivity. This causes the observed burnup enhancement of gas release. Using radial temperature profiles constructed from experimental data, grain growth and Xe diffusion during a power transient at high burnup are reexamined. It is found that grain growth is slower than predicted for steady-state irradiation conditions. The effective Xe diffusion coefficient, D Xe , in the transient tests with hold times of 40, 42 and 62 h is given by D Xe = 1.9 × 10 −8 exp (−21890/ T ) cm 2 s −1 , T in K. For the tests with a 4 h hold time, D Xe is giv by D Xe = 2.5 × 10 −9 exp (−21050/ T ) below 1570°C, but by the former equation above 1600°C. In the long transient tests, the increase in the D Xe below 1570°C is attributed to the annealing of gas traps in the oxide lattice.

Behaviour of Fission Gas in the Rim Region of High Burn-up UO2 Fuel Pellets with Particular Reference to Results from an XRF Investigation.

Abstract XRF and EPMA results for retained xenon from Battelles high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8–54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

Microstructure of irradiated SBR MOX fuel and its relationship to fission gas release

Abstract SEM and EPMA examinations of the microstructure and microchemistry of British Nuclear Fuel’s quasi-homogeneous SBR MOX fuel following irradiation suggests behaviour which is very similar to that observed in UO2. Most significantly, a fission gas release of 1% in three-cycle SBR MOX PWR rods is associated with the development of a well-defined intergranular bubble network, which has not been seen previously in the more heterogeneous MOX fuels irradiated under similar conditions. The contrast between the observations is attributed to the relatively low volume fraction and small size of the Pu rich inhomogeneities in the SBR fuel which generate only 4% of the total fission gas and eject most of this into the surrounding mixed oxide matrix. The resulting perturbation in the Xe distribution has a negligible influence on the evolution of the microstructure. A key observation is made from the results of recent post-irradiation annealing experiments performed on SBR MOX and UO2. These confirm near identical fission gas behaviour in the two fuel types when the influence of thermal conductivity and rod rating are removed.