High Burn-up Structure Research Articles

Effect of grain size on irradiation-induced bubble evolution in high burn-up UO2: A phase-field study

Due to economic reasons, uranium dioxide (UO2) ceramic fuel needs to operate under high burn-up conditions. The high burn-up structure (HBS) and micrometric irradiation-induced bubbles within it cause safety issues for the fuel. In this study, a phase-field model was developed to simulate the evolution of fission bubbles under high burn-up conditions. This model simultaneously couples the evolution of recrystallized grains and bubbles through explicit nucleation methods. By simulating the development of HBS in polycrystalline UO2, it was indicated that the microstructure is consistent with experimental observation. The variation of porosity was divided into three stages and found to be in good agreement with experimental data. Additionally, the HBS fraction was compared with experimental measurements to validate this HBS model. Furthermore, the effect of recrystallization on bubble evolution was investigated using the validated model. It was found that recrystallization accelerated bubble evolution by increasing GB area. In addition, the influence of grain size on bubble evolution was studied by constructing initial structures with different grain sizes. Results showed that both the development of recrystallization and the decrease in grain size enhance bubble evolution, but the enhancement becomes smaller as the grain size decreases.

Contributions to the mechanistic understanding of the microstructural evolution in irradiated U-Mo dispersion fuel

Advanced microstructural characterization techniques, such as scanning electron microscopy (SEM) and scanning transmission electron microscopy - energy dispersive x-ray spectroscopy (STEM-EDS), were used to interpret the fuel microstructure evolution and fission products behavior in U-Mo dispersion fuel irradiated in the Advanced Test Reactor (ATR) as part of the European Mini-Plate Irradiation Experiment (EMPIrE) test. The larger as-fabricated fuel grain size achieved by heat-treating the U-Mo powder resulted in slower high burnup structure (HBS) development and reduced fission gas porosity. Slower HBS kinetics was observed at the fuel kernels’ periphery, which contained smaller and less fission gas bubbles at all fission densities (FDs) investigated and was attributed to a locally reduced damage density and fission products concentration, as corroborated with Monte Carlo simulations. The non-refined grains at the fuel kernel periphery hosted a perfectly ordered fission Gas Bubble Superlattice (GBS) up to 6.3 × 1021 fissions/cm3. Nano-scale STEM-EDS analysis presented in this study provided useful information on the GBS characteristic morphology and evolution in U-Mo fuel. The concentration of fission gas in the GBS progressively increased with FD, pointing to an evolution of the nanobubble pressure status with irradiation. A possible connection between the GBS collapse and HBS onset is proposed for which there exists a threshold in the misorientation of the refined sub-grains above which the GBS stability during irradiation is no longer preserved, resulting in the GBS collapse.

STEM Analysis of High Burnup Structure in LWR Fuels.

STEM Analysis of High Burnup Structure in LWR Fuels.

Characterization of the radial microstructural evolution in LWR UO2 using electron backscatter diffraction

Studies on high burnup UO2 subjected to loss-of-coolant accident conditions have shown that restructured regions of the fuel are susceptible to pulverization and eventual dispersal. Due to a lack of pre-test characterization, the distinct microstructural features rendering the fuel prone to fragmentation remain ambiguous. Four samples of commercially irradiated light-water reactor UO2 have been characterized utilizing electron backscatter diffraction to assess the susceptible microstructure. The microscopy focused on determining the burnup and temperature conditions responsible for the formation of the different microstructural regions where the regions were denoted as the high-burnup structure (HBS), HBS transition, mid-radial, restructured central, and central region. Previous works have outlined the specific conditions required for the restructuring of the microstructure into the HBS, but the conditions responsible for the restructuring in the central region of the fuel are not well understood. The four analyzed samples confirm a burnup threshold of 61 GWd/tU, and an unknown temperature range is needed to facilitate the formation of the restructured central region. Additional fuel performance evaluations are needed to quantify the temperature range promoting restructuring in the central region.

A pore pressure model and cracking strength analysis on the UO2 high burnup structure

Advanced nuclear reactors are currently moving toward long fuel cycles and high burnup level. In the case of high burnup, plenty of fission gas will be produced, which promotes creation of overpressure pores and fracture in the fuel. The aim of this paper is to establish the cracking strength evaluation model of UO2 fuel and to obtain the stress field of high burnup UO2 fuel. And the non-uniform distribution of pore size, non-uniform distribution of pore location and the pore pressure model in rod-type dispersion fuel are considered. Firstly, Weibull distribution is used to describe the pore size distribution. With the aid of the relationship between local burnup and mean pore radius, geometrical models of the UO2 high burnup structure under different burnup conditions are established. Secondly, a pore pressure model taking into account pore size distribution, the relationship between local burnup and mean pore radius and Ronchi’s equation of state is established. Based on literature data, this work further verifies the rationality of using Weibull distribution to describe pore size distribution. Finally, the effects of pore distribution, local burnup and temperature on the pore pressure and maximum tension stress in the fuel are analyzed, and the stress-dominated pore coarsening mechanism is studied by finite element method. The conclusion can be drawn that the large size pores are mainly formed by the continuous coalescence of the small size pores with the surrounding pores. The fuel is not easy to crack when the pore size and location distribution is uniform. Both burnup and temperature increase contribute to fuel cracking.

The role of UC inclusions in the development of fission gas bubble superlattice neutron-irradiated monolithic U-10Mo fuels

Uranium Carbide (UC) inclusions are the most prevalent impurities in Uranium-Molybdenum (U-Mo) fuel and are considered undesirable because they could potentially affect fuel performance. This work revealed that, like grain boundaries, UC inclusions could help facilitate the formation of the fission gas bubble superlattice (GBS). The GBS is a highly organized complex defect structure that can effectively store fission gases, thereby inhibiting fuel swelling. Transmission electron microscopy (TEM) showed that GBS self-organization can initiate at the UC/U-Mo interfaces in U-10Mo fuel irradiated to low fission density. This study also revealed that the UC/U-Mo phase boundary in U-10Mo irradiated to low fission density is wavy and periodic in morphology and that GBS formation is semi-coherent with the UC boundary. Because the initiation of GBS preferentially occurs at grain boundaries and UC/U-Mo phase boundaries, it is expected that following GBS collapse at higher fission densities, high burnup structure development may initiate first at grain boundaries and UC/U-Mo interfaces.

Effect of burn-up on the thermal conductivity of light water reactor fuel: Results of investigations employing the laser flash technique

Analysing data drawn from three of EPRI's NFIR Programmes and the High Burn-up Rim Project (HBRP) managed by CRIEPI of Japan, expressions for the degradation of thermal conductivity with burn-up up to 103.5 MWd/kgHM are presented. These indicate that at high burn-up, above 70 MWd/kgHM, the thermal conductivity of the matrix of the high burn-up structure (HBS) increases by 25 - 30% concomitant with recrystallization of the fuel grains. This increase is partially or completely off-set, however, by a decrease in conductivity associated with the porosity of the HBS which contains the fission gas gathered from the fuel matrix during transformation of the microstructure. Below 70 MWd/kgHM, in the absence of the HBS, the conductivity is about 15% lower than that currently accepted. It is also found that the conductivity of UO2 fuel falls sharply immediately when irradiation begins. This is attributed to radiation damage rather than the incorporation of fission products in the fuel lattice. The data further reveal that at burn-ups below 70 MWd/kgHM the conductivity of UO2 containing 10 wt% Gd2O3 is markedly lower than that of UO2, but once the HBS has formed its conductivity is indistinguishable from that of conventional UO2, UO2 additive fuel or MOX. It is concluded that for accurate determination of the radial temperature profile in a high burn-up fuel pellet, knowledge of the volume fraction of the HBS in the pellet and the radial distribution of the porosity in the affected region is required.

Restructuring in high burn-up UO2 fuels: Experimental characterization by electron backscattered diffraction

This paper discusses the use of electron backscattered diffraction to characterize restructuring in a set of UO2 samples, irradiated in a pressurized water reactor at a burn-up between 35 and 73 GWd/tU, including standard UO2 samples and Cr-doped UO2 samples, to provide a better understanding of restructuring occurring both on the periphery and in the center of high-burn-up pellets. The formation of a high burn-up structure on the periphery of high burn-up UO2 was confirmed in our experiment. We found restructuring associated with bubble formation of all the samples in the central area, with higher irradiation temperatures when the burn-up exceeded 61 GWd/tU, regardless of their initial microstructure. This restructuring tended to progress with the increasing burn-up and to sub-divide the initial grains into sub-grains, with orientations close to that of the parent grains. Radial changes and differences between these samples showed that the burn-up and the temperature were not the only relevant parameters involved in restructuring.

Investigation on axial relocation of fragmented annular fuel under loss of coolant accident conditions with discrete element method

Axial fuel relocation of conventional cylindrical fuel rods under Loss of Coolant Accident (LOCA) conditions has been intensively studied. However, characteristics of cladding ballooning, fuel fragmentation and relocation of a dual-cooled annular fuel rod could be quite different due to its innovative geometry. In this study, based on fragmentation images of annular fuels after in-pile irradiation, discrete geometry for fuel fragments was established considering the effect of burnup. With the ballooning cladding as constraint boundary and the gravity as driving force, axial relocation of an annular fuel at different burnup stages was simulated with discrete element method (DEM). The employed DEM methodology was first validated through a benchmark against a theoretical model and a reference simulation. Simulation results indicate that relocation is determined by the fuel fragmentation morphology and cladding ballooning shape. At low and medium burnup stages, it was found that although there exists obvious mass accumulation in the ballooning region, many fuel fragments above the ballooning region are stuck between the internal and external claddings, resulting in almost no missing length of the fuel stack. At high burnup stages, because of the presence of finer fuel particles at High Burnup Structure (HBS) regions, fuel jam can be mitigated or avoided, resulting in a higher peak mass fraction and an obvious missing fuel length. The critical HBS volume fraction that can totally avoid the fuel jam was determined as about 15%. In addition, the sensitivity study indicates that the higher the volume proportion of HBS is, the more obvious the axial fuel relocation is. In the absence of sufficient irradiation data, this study provides a reliable modeling methodology and valuable understandings of the axial fuel relocation in dual-cooled annular fuel rods.

Observations of high burnup structure in AGR fuel

Nuclear power plants (NPPs) within the UK are largely of the commercial advanced gas-cooled reactor (AGR) design. The key elements of the reactor and fuel design differ somewhat to those seen in Light Water Reactors (LWRs), most markedly the cladding, an austenitic stainless steel, and annular ceramic UO2 fuel pellets. The graphite moderated AGR cores also experience much higher temperatures than PWR (Pressurised water reactor) cores.In order to fully understand the requirements for enhanced fuel performance and interim/long term spent fuel storage, some efforts must be made to compare and contrast the two fuel types, to understand if LWR data are fully applicable to AGR data in various operational and storage scenarios. One such microstructural effect requiring assessment is a fuel microstructure commonly referred to as high burnup structure (HBS).The current study describes a series of post-irradiation examinations (PIE) undertaken on AGR fuel. A total of circa 30 AGR samples are examined in which HBS is present, and a further similar size of samples in which no HBS was present. Observations include SEM (scanning electron microscopy) and LOM (light optical microscopy) providing a characterisation of the appearance of HBS in AGR. The results are supported by burnup profile modelling, and assessment of fuel temperature from metallographic markers.The results are consistent with AGR outer ring fuel exhibiting an onset of HBS at a rod average burnup of ∼30 MWd/kgU and a 100 μm HBS thickness at a rod average burnup of 40 MWd/kgU. These values are lower than the comparable (pellet average) burnup for PWR fuel. The difference in burnup is shown to be due to the neutron spectrum in AGR fuel elements, and which also leads to a marked variation in HBS thickness around the circumference of AGR fuel. There is no suggestion that the local onset burnup for HBS formation differs between the two fuel types.

Effects of fuel relocation on fuel performance and evaluation of safety margin to 10CFR50.46c ECCS acceptance criteria in APR1400 plant

Characteristics of fuel relocation and its effect on safety margin in a loss-of-coolant accident (LOCA) safety analysis have been analyzed. Packing fraction model developed by Quantum Technology was validated and slight modification was conducted based on the formation of high burnup structure model. FAMILY computer code that is an integrated code with FRAPTRAN and MARS-KS is used for the evaluation. The peak cladding temperature (PCT) and peak local oxidation (PLO) of the APR1400 nuclear power plant with the PLUS7 fuel during LOCA were evaluated by a best-estimate plus uncertainty analysis methodology up to 60 MWd/kgU. Fuel relocation resulted in an increase in 95 %/95 % probability/confidence level of the reflood PCT (95/95 reflood PCT) and the PLO (95/95 PLO) until the fuel burnup of ∼ 50 MWd/kgU. Maximum increase of 95/95 reflood PCT was about 130 K at 30 MWd/kgU burnup. This results in the minimum safety margin of the reflood PCT at that burnup with respect to the U.S. NRC proposed 10CFR50.46c PCT limit. Maximum increase of 95/95 PLO was 3.9 percent point at 30 MWd/kgU. However, minimum safety margin was observed at different burnups such as 40 MWd/kgU and 60 MWD/kgU. Regarding the cladding burst characteristics, fuel relocation may lead to an increase in the burst temperature and burst strain of the cladding, and it also induces an earlier failure of cladding.

A phase-field model of quasi-brittle fracture for pressurized cracks: Application to UO[formula omitted] high-burnup microstructure fragmentation

In this paper, we present a phase-field model of quasi-brittle fracture with pressurized cracks, with dedicated applications for polycrystalline materials. The model is formulated as a minimization problem within the variational framework. The external work done by pressure on the crack surfaces is included in the objective function. Several careful modeling choices lead to a regularization-length-independent critical strength. The model is constructed to give a softening response with an underlying linear traction-separation law. The pressure-dependent softening response and the regularization of the prescribed pressure are demonstrated with a (quasi) one-dimensional numerical analysis. In a two-dimensional numerical analysis under plane strain assumptions, the critical stress corresponding to crack propagation (as predicted by our quasi-brittle fracture model) is compared with linear elastic fracture mechanics (LEFM) analytical solutions. Our model is further utilized to simulate fission-gas-induced fragmentation of the UO2 high-burnup structure (HBS). Simulation results show that pressurized bubbles can cause crack nucleation and propagation, and that the bubble size and the surrounding external pressure affect the critical pressure corresponding to crack nucleation. Simulations of a partial HBS at different recrystallization stages show that different grain structures (due to recrystallization) also influence crack paths and fragmentation morphology.

Molecular dynamics studies of grain boundary mobility and anisotropy in BCC γ-uranium

Grain morphologies such as grain size and aspect ratio in uranium-based metallic fuels are important microstructural features that can impact various fuel performance properties such as fission-gas-induced swelling, thermal transport, high burnup structure formation, and radiation resistance. Accurate prediction of the fuel grain morphologies requires knowledge of critical grain growth parameters such as grain boundary (GB) mobility and anisotropy. In this work, molecular dynamics (MD) simulations were performed to study the GB mobility and its anisotropy in pure body-centered-cubic (BCC) γ uranium. Nine GBs with different combinations of misorientation angles (20°, 30°, 45°) and rotation axes (<100>, <110>, <111>), as well as an additional <111> 38.2° GB were studied using three interatomic potentials. It is found that the GB mobility anisotropy has complex trends, depending on both rotation axis and misorientation. However, in general the <110> rotation axis has the fastest GB mobility at the same misorientation. The results of this work can be used as not only a baseline for future studies of GB mobility in uranium-based alloys such as uranium-molybdenum (U-Mo) fuels, but also input for mesoscale modeling of grain growth in uranium-based alloys.

Identifying chemically similar multiphase nanoprecipitates in compositionally complex non-equilibrium oxides via machine learning

Characterizing oxide nuclear fuels is difficult due to complex fission products, which result from time-evolving system chemistry and extreme operating environments. Here, we report a machine learning-enhanced approach that accelerates the characterization of spent nuclear fuels and improves the accuracy of identifying nanophase fission products and bubbles. We apply this approach to commercial, high-burnup, irradiated light-water reactor fuels, demonstrating relationships between fission product precipitates and gases. We also gain understanding of the fission versus decay pathways of precipitates across the radius of a fuel pellet. An algorithm is provided for quantifying the chemical segregation of the fission products with respect to the high-burnup structure, which enhances our ability to process large amounts of microscopy data, including approaching the atomistic-scale. This may provide a faster route for achieving physics-based fuel performance modeling.

Application of BISON to UO[formula omitted] MiniFuel fission gas release analysis

There has been a recent push to accelerate nuclear fuel qualification by combining advanced modeling and simulation with accelerated separate effects irradiation testing. One separate effects irradiation testing capability is the MiniFuel vehicle designed to accelerate burnup accumulation in “mini” fuel samples under isothermal temperature conditions. These steady-state MiniFuel irradiations effectively decouple the fuel temperature from the fission rate (i.e., power) by minimizing the fuel volume and relying on gamma heating in the surrounding components for temperature control. This provides experimenters a flexible means for targeting and evaluating specific fuel microstructures for a wide range of fuel types and operating conditions. However, the accelerated fuel qualification process must be informed by modeling and simulation to properly evaluate the most impactful fuel performance parameters and to identify modeling and data gaps. A test matrix can then be designed to fill those gaps and, ultimately, refine and validate the fuel performance models. This work uses the BISON fuel performance code to conduct an assessment and sensitivity study on calculated fission gas release (FGR) from UO2 MiniFuel disks during isothermal irradiation and temperature transients. Qualitatively, the existing fission gas behavior model in BISON reproduces the effects of temperature and burnup as expected. However, when quantitatively compared with FGR data from irradiated (103 MWd/kgU) UO2 disks under thermal annealing representative of LOCA conditions, the model shows a less satisfactory predictive capability. As a result, development needed to model the specific mechanisms of transient FGR during LOCAs, including fuel fragmentation and the role of the high burnup structure (HBS), are identified. Finally, a UO2 MiniFuel test matrix is proposed to extend the temperature and burnup ranges covered by previous experiments and provide new model validation data for fission gas behavior under high burnup and transient conditions.

Modeling high burnup structure in oxide fuels for application to fuel performance codes. Part II: Porosity evolution

We propose a model describing the high burnup structure inter-granular porosity evolution under irradiation. The evolution of the porosity collecting the gas diffusing from the grains is modeled by exploiting a second-order Fokker-Planck expansion of the cluster-dynamics master equations governing the problem, considering nucleation of pores, gas absorption due to the diffusional flow from the grains, size-dependent re-solution of gas from pores due to interaction with fission fragments, vacancy absorption, and pore coalescence. Model predictions on xenon local retention, matrix fuel swelling, and porosity evolution are compared to experimental data and to models available in fuel performance codes.

High burnup structure formation in U-Mo fuels

This study used electron microscopy techniques and phase field modeling to investigate the mechanisms responsible for developing the high burnup structure in U-Mo fuels. The results show that grain subdivision is initiated by polygonization; however, evidence of dynamic recrystallization is present with increasing fission densities. At 2.5 × 1021 fissions/cm3, LAGB misorientation of 4° was measured near the grain boundary of neighboring grains, which supports polygonization as the leading grain subdivision mechanism. However, with increasing fission densities, the formation of HAGB subgrain clusters are evident, marking the activation of another mechanism − dynamic recrystallization. Previously formed LAGBs transform into new HAGB grains, which is reflected in the increasing number frequency of HAGBs to LAGBs. At the same time, the HAGB subgrain clusters also undergo polygonization, suggesting that polygonization is a continuous mechanism. Based on the simulations performed in this study, polygonization may continue to occur until a mean grain diameter of ∼382 nm is achieved. Based on the results of this work, LAGB to HAGB transformation is associated with a strain-driven mechanism potentially caused by fission gas-induced plastic deformation and is consistent with recrystallization. Polygonization and dynamic recrystallization occurs in tandem with polygonization leading grain subdivision. Based on the fission densities evaluated in this study, authors postulate that the critical fission density at which dynamic recrystallization (DRX) becomes active is between 2.5 × 1021 fissions/cm3 and 3.5 × 1021 fissions/cm3.

Oxidation of Micro- and Nanograined UO2 Pellets by In Situ Synchrotron X-ray Diffraction.

When in contact with oxidizing media, UO2 pellets used as nuclear fuel may transform into U4O9, U3O7, and U3O8. The latter starts forming by stress-induced phase transformation only upon cracking of the pristine U3O7 and is associated with a 36% volumetric expansion with respect to the initial UO2. This may pose a safety issue for spent nuclear fuel (SNF) management as it could imply a confinement failure and hence dispersion of radionuclides within the environment. In this work, UO2 with different grain sizes (representative of the grain size in different radial positions in the SNF) was oxidized in air at 300 °C, and the oxidation mechanisms were investigated using in situ synchrotron X-ray diffraction. The formation of U3O8 was detected only in UO2 pellets with larger grains (3.08 ± 0.06 μm and 478 ± 17 nm), while U3O8 did not develop in sintered UO2 with a grain size of 163 ± 9 nm. This result shows that, in dense materials, a sufficiently fine microstructure inhibits both the cracking of U3O7 and the subsequent formation of U3O8. Hence, the nanostructure prevents the material from undergoing significant volumetric expansion. Considering that the peripheral region of SNF is constituted by the high burnup structure, characterized by 100–300 nm-sized grains and micrometric porosity, these findings are relevant for a better understanding of the spent nuclear fuel behavior and hence for the safety of the nuclear waste storage.

Study of the hardness and Young's modulus at the fuel-cladding interface of a high-burnup PWR fuel rod by nanoindentation measurements

During irradiation in a nuclear reactor, the fuel swelling and the cladding creep are at the origin of the contact between the fuel and the cladding, which in turn leads to the initiation of Zr oxidation by the UO2 fuel. In high burnup fuels, the zirconia layer formed at the Fuel Cladding Interface (FCI) presents numerous circumvolutions and a heterogeneous microstructure within its thickness. In this paper, nanoindentation measurements are carried out in the fuel-cladding interface of a PWR fuel rod irradiated up to a high burnup of 61 GWd/tU. The measured hardnesses and Young's moduli of the internal zirconia layer are compared to reference measurements obtained on a non-irradiated sample of commercial Y-TZP. The study of the evolution of mechanical properties across the zirconia layer led to the identification of four distinct zones related to the zirconia microstructure (phase, grain size, porosity and fission product recoil). Close to the cladding, the zirconia hardness HZrO2tends toward the zirconium hardnessHZr. Then, the zirconia hardness increases significantly with the grain size, until reaching a plateau at the middle of the zirconia layer, where the maximum hardness is obtained. Following the same trend, the elastic modulus progressively increases in the zirconia layer and reaches a plateau near the fuel High Burnup Structure.

Iodine release from high-burnup fuel structures: Separate-effect tests and simulated fuel pellets for better understanding of iodine behaviour in nuclear fuels

Iodine release modelling of nuclear fuel pellets has major uncertainties that restrict applications in current fuel performance codes. The uncertainties origin from both the chemical behaviour of iodine in the fuel pellet and the release of different chemical species. The structure of nuclear fuel pellet evolves due to neutron and fission product irradiation, thermo-mechanical loads and fission product chemical interactions. This causes extra challenges for the fuel behaviour modelling. After sufficient amount of irradiation, a new type of structure starts forming at the cylindrical pellet outer edge. The porous structure is called high-burnup structure or rim structure. The effects of high-burnup structure on fuel behaviour become more pronounced with increasing burnup. As the phenomena in the nuclear fuel pellet are diverse, experiments with simulated fuel pellets can help in understanding and limiting the problem at hand. As fission gas or iodine release behaviour from high-burnup structure is not fully understood, the current preliminary study focuses on (i) sintering of porous fuel samples with Cs and I, (ii) measurements of released species during the annealing experiments and (iii) interpretation of the iodine release results with the scope of current fission gas release models.Graphical abstract