Large Fission Gas Research Articles

Oriented intra-granular bubble transport due to coupling of pinned bubble growth and dislocation climb

The intra-granular fission gas release during post-irradiation annealing tests cannot be predicted if bubble motion is not taken into account. Indeed, without irradiation, the equilibrium between trapping and re-solution is completely shifted in favor of trapping, the dynamic resolution due to fission spikes no longer existing. Moreover, the sink strength of the population of intragranular bubbles is greater by a factor ∼1000 than the sink strength of the grain surface. Trapped in supposedly immobile bubbles, the gas would not escape from the grain, which is in contradiction with experimental observations. Several alternative scenarios involving bubble movement emerged to explain the observed fission gas release. The purpose of our work is to assess these scenarios using simulation. In a previous article, it was demonstrated that neither the movement of bubbles in a vacancy gradient, nor the Brownian movement of bubbles, nor the combination of them, could explain the large fission gas release obtained during post-irradiation annealing in our reference experiment. This demonstration was performed using a mesoscale model, called BEEP, where individual bubbles are described, along with the diffusion of vacancies from each bubble to the other, as well as from the free surface. In this paper, we extend the BEEP model to assess the role of dislocations in interaction with highly pressurized bubbles. It is concluded that a mechanism of dislocation climb coupled with the growth of highly pressurized pinned bubbles may explain the large intra-granular fission gas release in annealing conditions.

Possible impacts of Mo chemical banding and second phase impurities on the irradiation behavior of monolithic U-10Mo fuels

This study investigated the microstructural behavior of both full-size and mini-size monolithic U-10Mo fuel plates irradiated to high burnup with a focus on the evolution of the second phase impurities in monolithic U-Mo using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and wavelength dispersive spectroscopy (WDS). Key indicators of possible mechanical and thermal compromise include cracks, large fission gas porosity, and interconnection of fission gas pores. For the fission densities evaluated in this work (3.5 × 1021 fissions/cm3 -5.1 × 1021 fissions/cm3), fine porosity can develop along the UC phase boundary; however, the size of the fission gas pores is no more than those observed in the U-Mo fuel phase. Other inclusions such as Si-rich second-phase impurities found in the as-fabricated microstructure were difficult to resolve post-irradiation because they can become overshadowed by porosity development in the fuel phase. Additionally, the presence of a Fe-rich sublayer formed in the Zr diffusion layer during fabrication remained enriched in the Zr layer in the irradiated U-10Mo microstructures near the U-Mo/Zr interface; however, based on the burnup assessed in this study the identified impurities did not appear to contribute to notable microstructural degradation.

An investigation of the failure modes in U-10Mo monolithic fuel irradiated to high burnup

This study investigated possible failure mode sequences in U-10Mo monolithic fuel irradiated to very-high-burnup. The U-Mo fuel plate used in this study was characterized using scanning electron microscopy (SEM), wavelength dispersive spectroscopy (WDS), and image analysis techniques to investigate how microstructural features may initiate crack formation and propagation through the microstructure, potentially leading to blistering. Distinctly large fission gas pores (FGPs) were observed to preferentially align 5-15µm away from the U-10Mo/Zr interface, parallel to the Zr diffusion barrier. Preferential growth, alignment, and interconnection of FGP could initiate blistering in the fuel plate by creating a fission gas channel, parallel to the U-Mo/Zr interface. FGP alignment in the fuel phase near the U-Mo/Zr interface is believed to be one of the precursors to crack and Type 2 blister formation in monolithic U-Mo fuels. The chemical maps revealed differences in fission product behavior near the U-Mo/Zr interface such that Nd remains in the fuel matrix at the FGP sites. On the other hand, other fission products like Xe and Cs can get trapped in FGPs by Nd in the fuel phase and/or diffuse toward the Zr diffusion barrier.

Research on performance enhancement of nuclear fuel with SiC cladding by using high thermal conductivity fuels

The engineering application of the fuel rod system consisting of traditional UO2 pellets and SiC accident tolerant fuel (ATF) cladding is challenged by high fuel operating temperatures, large fission gas release rates and high risks of brittle failure of the cladding. In this study, a preliminary research was carried out for performance enhancement of the fuel rods consisting of SiC cladding by replacing UO2 with high thermal conductivity fuels, including UO2/BeO, U3Si2 and UN/U3Si2. Thermo-mechanical modeling was conducted with the consideration of burnup effects for the three fuels with high thermal conductivity. An upgrade of the nuclear fuel analysis code FRAPCON4.0 was conducted to extend the simulation capability by implementing the new fuel models. Thermo-mechanical behavior of the original and improved fuel rods was simulated by the upgraded FRAPCON4.0. The simulation results show that using high thermal conductivity fuels can significantly reduce the fuel temperature and fission gas release. Furthermore, the fuel gaseous swelling is greatly mitigated due to the reduced fuel temperature, which is beneficial to prevent SiC cladding from brittle failure caused by pellet-cladding mechanical interaction. By comparison, UN/U3Si2 fuel presents the most promising and desired thermo-mechanical performance among all high thermal conductivity fuels investigated in this study.

Modelling intra-granular bubble movement and fission gas release during post-irradiation annealing of UO2 using a meso-scale and spatialized approach

The effective diffusion theory, which is generally used in the modelling of base-irradiation of nuclear fuel, cannot predict the intra-granular fission gas release during post-irradiation annealing tests. From this discrepancy between experiments and usual theory, several alternative scenarios emerged. The purpose of this work is to model these scenarios, as mechanistically as possible, and to distinguish those that could really explain the observations. The difficulty is that the fission gas bubbles in irradiated UO2 are extremely small and numerous (mean distance between nano-bubble centers is only of the order of 10–20 nm) while the grain radius is about 5μm. A new spatialized mesoscale model was developed where individual bubbles are described, along with the diffusion of vacancies from each bubble to the other, as well as from the free surface. Random movement and coalescence of the bubbles have also been included in the model. Based on this principal, two scenarios, and the combination of those, could be assessed: (a) the movement of bubbles in a vacancy gradient, and (b) the Brownian movement of bubbles. It was demonstrated that neither of these two scenarios, nor the combination of them, could explain the large fission gas release obtained during post-irradiation annealing in our reference experiment. This encourages us to consider additional mechanisms, involving dislocations for instance, that could explain the high fission gas release.

The use of U3Si2/Al dispersion fuel for high power research reactors

In order to investigate the irradiation performance of U3Si2/Al dispersion fuel at relatively aggressive reactor conditions, compared to how this fuel is typically used in a research or test reactor, post irradiation examination was performed on two fuel plates that were irradiated at ∼270 W/cm2 average surface heat flux and to a maximum local burnup greater than 8 × 1021 fissions/cm3 in the Advanced Test Reactor as part of the RERTR-8 experiment. As part of the non-destructive examinations that were performed on both fuel plates, detailed thickness measurements using a high-fidelity measurement bench were performed on one fuel plate and a segment of the second fuel plate. This paper provides the first reported results of such high-resolution thickness measurements for U3Si2/Al dispersion fuel plates, containing primarily the U3Si2 fuel phase, irradiated at such power levels to high burnup. For destructive examination using optical metallography, samples were generated from both irradiated plates. Localized regions of high swelling (characteristic of internal blistering) were observed on the surface of both fuel plates, suggesting that the performance limits had been reached for each plate. Destructive examinations showed that the fuel particles that are primarily U3Si2, which become amorphous during irradiation, in both plates contained relatively large fission gas bubbles with evidence of fission gas bubble interconnection.

Modeling of Pore Coarsening in the Rim Region of High Burn-up UO 2 Fuel

Abstract An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated UO 2 fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO 2 fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO 2 fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

Performance of low smeared density sodium-cooled fast reactor metal fuel

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

Temperature and dose dependence of fission-gas-bubble swelling in U 3Si 2

Large fission gas bubbles were observed during metallographic examination of an irradiated U 3Si 2 dispersion fuel plate (U0R040) in the Advanced Test Reactor (ATR). The fuel temperature of this plate was higher than for most of the previous silicide-fuel tests where much smaller bubble growth was observed. The apparent conditions for the large bubble growth are high fission density (6.1 × 10 21 f/cm 3) and high fuel temperature (life-average 160 °C). After analysis of PIE results of U0R040 and previous ANL test plates, a modification to the existing athermal bubble growth model appears to be necessary for high temperature application (above 130 °C). A detailed analysis was performed using a model for the irradiation-induced viscosity of binary alloys to explain the effect of the increased fuel temperature. Threshold curves are proposed in terms of fuel temperature and fission density above which formation and interconnection of bubbles larger than 5 μ are possible.

Low-temperature irradiation behavior of uranium–molybdenum alloy dispersion fuel

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium–molybdenum (U–Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4–10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel–matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U–10Mo composition. Both of the U–10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

The irradiation performance of magnesia-doped UO 2 fuel

Large-grained UO 2 doped with 5 mol% magnesia has been irradiated in the Windscale Advanced Gas Cooled Reactor to 28 500 MWd/ te( U). Fission gas release from the doped fuel was reduced by a factor greater than 2.5 × in comparison to that from standard UO 2 irradiated under identical conditions. Micro-gamma scanning showed that the retention of 137CS was also considerably greater in the doped fuel. Although computer calculations suggest that the improved fission product retention at this burn-up can be accounted for by differences in grain size, microstructural examination of the irradiated fuel showed considerable differences in the morphology of fission gas precipitation between the doped and undoped fuels. In particular, large fission gas bubbles ~0.2 μm in diameter were observed within the grains of doped fuel. It is suggested that these are stabilised by interaction with the MgO precipitates and that this may reduce diffusion of gas to the grain boundaries at higher burn-ups.