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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Characterization of high thermal conductivity fuel surrogates before and after ion irradiation

High thermal conductivity nuclear fuels offer important potential advantages over traditional oxide-based fuels such as higher burnup, reduction in fission gas release, and better overall safety of the system. One proposed approach to high thermal conductivity fuels utilizes high thermal conductivity nonfissile additives with UO 2 fuel to lower the fuel operating temperature and thereby take advantage of the highly favorable radiation resistance of UO 2 at lower operating temperatures. However, differential swelling in the matrix and high conductivity additive phases during high dose irradiation could lead to internal cracking and poor performance. In the current study, ceria (CeO 2 ) and zirconia (ZrO 2 ) surrogate matrices were used to model UO2 behavior. Additives of 10 vol. % Al 2 O 3 or SiC in the form of short fibers or platelets were used for the high conductivity second phase. The nuclear fuel surrogates were sintered to achieve densities greater than 93% of the ideal values. Scanning electron microscopy (SEM) imaging and X-ray diffraction confirmed the uniform distribution of the second phase and that no intermetallic second phase was formed during sintering. The thermal conductivity of the sintered samples was measured from 50 °C to 900 °C and confirmed the desirable increase compared to pure CeO 2 /ZrO 2 pellets. Samples were irradiated with 20 MeV Ni 6+ ions at midrange doses ranging from 1 to 15 displacements per atom (dpa) and temperatures from 300 °C to 700 °C. Post irradiation characterization revealed a good stability of the samples at low to medium doses with matrix lattice parameter swelling of < 0.14 % but showed a significant microstructural deterioration and decrease of the mechanical properties at 15 dpa.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Thermochemically-informed mass transport model for interdiffusion of U and Zr in irradiated U-Pu-Zr fuel with fission products

A new formulation for redistribution of constituents in metallic U-Pu-Zr nuclear fuel is presented that can incorporate the contributions from fission products. The formulation is based on the thermodynamic driving forces derived from the generalized chemical potential that includes effects of composition and temperature. As a result, the redistribution model can readily account for the composition changes due to the generation of fission products while using only a limited set of transport coefficients. The thermodynamic model for the metallic fuel and thermochemistry solver Thermochimica were coupled with the nuclear fuel performance code BISON to implement the redistribution model. The simulations reproduce an experimentally observed Zr-depletion zone in the mid-radius region of an irradiated fuel slug. The generation of fission products during burnup is shown to have a stabilizing effect on the fuel chemistry, slowing the rate of U-Zr interdiffusion and reducing the size of the Zr-depleted zone. Here, the overall objective of the work is to advance predictive capabilities of fission product behaviour in U-Zr metallic fuel in the context of fuel performance and safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Improvement of the $\mathrm{BISON U_3Si_2}$ modeling capabilities based on multiscale developments to modeling fission gas behavior

Uranium silicide (U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel, but operated at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. These lower length scale informed models are then utilized in the validation of BISON U 3 Si 2 modeling capabilities to simulate the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the validation process to identify where further experiments and lower length scale modeling would be beneficial. Here, the multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

First-principles study of the surface properties of LiAl 5 O 8 : Stability and tritiated water formation

LiAl 5 O 8 is an important secondary phase of γ-LiAlO 2 ceramics in tritium ( 3 H, T)-producing burnable absorber rods (TPBARs), present in small quantities in the as-fabricated condition and in larger quantities after irradiation due to Li burnup and mobility in the radiation-damaged lattice. In this work, we performed first-principles calculations to study the stabilities and structures of LiAl 5 O 8 surfaces the possibility of T 2 O formation on the (111) surface. By calculating the surface energies of all possible symmetrical stoichiometric and nonstoichiometric low-index surfaces, we identified the two most stable surfaces, one stoichiometric (001) surface and one nonstoichiometric (111) surface in an O-rich condition. The surface oxygen atoms on the nonstoichiometric (111) surface have remarkable displacements after relaxation, due to the large dangling O bonds. Considering a Li-poor and 3 H-rich condition relevant to TPBARs, we further studied these two stable surfaces with 37.5%, 75%, and 100% surface O atoms covered by 3 H atoms. Our results show that the 3H saturation is favorable for the nonstoichiometric (111) surface but not for the stoichiometric (001) surface. In addition, the structure of nonstoichiometric (111) surface is well stabilized when the surface O atoms are 100% saturated by 3 H atoms. Based on the fully 3H-saturated (111) surface, the 3 H 2 O desorption with an energy barrier of 0.69 eV is preferable in the 3 H-rich condition.

36 MATERIALS SCIENCE↗
DOE OSTI2021

25-Pin metallic fuel performance benchmark case based on the EBR-II X430 experiments series

A metallic fuel benchmark case was developed for the fuel performance code BISON based on 25 uranium-zirconium and uranium-plutonium-zirconium pins of the Experimental Breeder Reactor II X430 experiment series. Results of the benchmarks were compared with measurements and calculations made at the time of the experiment as well as subsequent measurements reported in 2019. The comparisons were used to quantify the accuracy of the BISON predictions and to identify patterns in the BISON differences.BISON predicted burnup, plenum pressure, and fission gas release fractions accurately. BISON temperature predictions were somewhat cooler than the temperatures determined at the time of the experiment but appeared to be reasonably accurate considering uncertainties in the legacy temperature calculations, uncertainties in the legacy linear heat rate calculations, and the high sensitivities of the BISON-predicted temperatures to BISON inputs. Fuel axial elongation predictions had errors correlated to fuel composition; BISON tended to underpredict the elongation of binary fuels and overpredict the elongation of ternary fuels. BISON cladding radial dilation predictions were also significantly lower than legacy PIE measurements.Recommendations were made to improve the BISON fuel gaseous swelling model to account for fuel composition, to add additional capabilities to the coolant channel temperature model to ease benchmark development, and to continue developing benchmark cases based on a wide range of experiments in several reactors. Once a wide array of benchmarks is developed, an attempt can be made to enhance or calibrate BISON models to improve the cladding dilation predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Investigation of constituent redistribution in U-Pu-Zr fuels and its dependence on varying Zr content

This contribution investigates fuel constituent segregation and fuel-cladding chemical interactions (FCCI) in three U-Pu-Zr fuel pins irradiated to ~11% burnup in Experimental Breeder Reactor-II as part of the X441 DP-1 experiment. In examined pins, Zr content ranged from 6 to 10 and 14 wt.%, while Pu concentration was constant at 19 wt.% Pu. The primary goal of the investigation was to determine the role Zr content plays in fuel performance and this manuscript provides assessment of both constituent redistribution and FCCI as a function of axial position. FCCI was observed in all pins, though no cladding failure was noted. Optical and scanning electron microscopy (SEM) results show that variation in Zr content alters the number and relative size of constituent redistribution zones in the fuel. For example, larger Zr-rich central regions and smaller U-rich intermediate regions were observed in 14 wt.% Zr fuels. More importantly, all examined fuels have four or more discrete constituent redistribution regions, where each region has dissimilar morphological features. The existence of four to six distinctive regions deviates from traditionally accepted three region redistribution model, highlights complexity of constituent redistribution in this metal fuel, and identifies the need to conduct additional studies using state-of-the-art instrumentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Postirradiation characterization of palladium as an additive for fuel cladding chemical interaction mitigation in metallic fuel

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Furthermore, the efficacy of using this additive needs further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Application of BISON to UO 2 MiniFuel fission gas release analysis

There has been a recent push to accelerate fuel qualification by developing revolutionary capabilities to reduce irradiation periods, and thereby, reduce the time required to qualify a new fuel system. One such capability is the MiniFuel irradiation capsule designed to miniaturize fuel samples and irradiate “mini” fuel samples under isothermal temperature conditions. MiniFuel allows steady-state irradiations to decouple the traditionally coupled fission rate (i.e., power) and temperature parameters to understand and generate microstructures observed in fuel operated in a commercial reactor. Furthermore, this process offers the possibility to gather in situ data as well as postirradiation or transient data such as thermal conductivity, specific heat, fission gas diffusion and release, etc. However, accelerating fuel qualification is not solely reliant on generating large amounts of data but also on developing an informed test matrix designed to rapidly generate impactful data. Additionally, this process is reliant on fuel performance codes, such as BISON, to evaluate MiniFuel irradiations using existing material models. This process pinpoints model/data gaps, identifies desired irradiation conditions, and subsequently supports model validation and development. This work describes the use of BISON to perform a number of sensitivity studies designed to understand conditions that lead to fission gas release (FGR) under steady-state isothermal irradiation conditions and temperature transient conditions. The model is applied to a UO 2 MiniFuel example and shows an overall good qualitative agreement with experimental FGR annealing tests under different temperature conditions. It also accounts well for microstructural effects on FGR. When quantitatively compared with FGR data from previously irradiated 103 MWd/kgU UO 2 discs under thermal annealing, the model shows a less satisfactory agreement with the experimental data. Finally, a UO 2 MiniFuel test matrix is proposed to help to extend the model's operational range and validate the new FGR model capabilities to higher burnups and transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A study on texture stability and the biaxial creep behavior of as-hydrided CWSR Zircaloy-4 cladding at the effective stresses from 55 MPa to 65 MPa and temperatures from 300°C to 400°C

The creep rupture of high burnup used nuclear fuel (> 45 GWD/MTU) cladding is regarded as one of the failure mechanisms during long-term dry storage. A high amount of zirconium hydride in the cladding matrix would degrade the mechanical properties of the cladding especially leading to delayed hydride cracking. To better understand the influence of zirconium hydride on the biaxial thermal creep performance and the crystalline texture stability of Zircaloy-4 cladding, the pressurized tube technique is applied to test the durability of the as-hydrided material. Tests were performed on as-received Zircaloy-4 tubular specimens as well as-hydrided ones with targeted 300 wt parts per million (wppm) or 750 wppm hydrogen. The biaxial creep experiments were conducted at temperatures from 300 degrees C to 400 degrees C and at equivalent stresses at mid-wall from 55 MPa to 65 MPa. The hydridation process prior to creep tests induces the formation of FCC delta-hydride platelets along the circumferential direction of the tube. This alignment and phase structure of hydride show no significant change after biaxial creep tests. The creep strain-rate negatively depends on the hydrogen content. The synchrotron wide-angle X-ray diffraction (WAXD) technique and electron backscatter diffraction (EBSD) analysis were applied for the study of the crystallographic orientation relationship. zirconium-to-hydride grain orientation follows Shoji-Nishiyama crystallographic relationship. This relationship is stable before and after creep deformation. These results of creep performance and texture stability of Zircaloy-4 claddings can help support the design basis of interim and long-term dry storage facilities.

creep↗
DOE OSTI2022

Anisotropic temperature-dependent elastic constants and thermal conductivities of TRISO particle coatings

Tristructural isotropic (TRISO) particles show great promise as a candidate fuel for use in several next-generation high-temperature nuclear reactor designs due to their structural integrity and fuel performance at high temperatures and burnups. Computational codes exist that can simulate TRISO fuel performance characteristics and failure probabilities under extreme conditions which require knowledge of the TRISO coatings’ thermophysical properties. The thermophysical descriptions of the TRISO particle's layers (i.e., buffer, pyrolytic carbon, and silicon carbide) currently used in fuel performance codes, however, assume that many of these properties are constant with respect to temperature or texture. In this paper, we obtain the full elastic stiffness tensors of the carbon and silicon carbide layers, which have transversely isotropic symmetry. Additionally, the calculated elastic properties of each of the coating layers are in remarkable agreement with the current models at room temperature and correct orientations. Additionally, the textured 3C-SiC layer was found to exhibit novel auxetic behavior above 1500 °C. The anisotropic temperature-dependent thermal conductivities of the carbon layers were calculated using acoustical Grüneisen-Debye theory which are in excellent agreement with current models at room temperature and correct orientations. These texture- and temperature-dependent relationships can be incorporated into the thermophysical description of TRISO particles in order to more accurately model fuel performance and failure probabilities under extreme conditions in forthcoming high-fidelity computational simulations.

36 MATERIALS SCIENCE↗
DOE OSTI2022

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 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. Here, 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 baselines 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Transmission Electron Microscopy based Characterization of a U-20Pu-10Zr Fuel Irradiated in Experimental Breeder Reactor-II

U-Pu-Zr metallic fuels are important fuel candidate for future advanced and/or test reactors. To better understand fuel performance and guide future fuel development, a U-20Pu-10Zr (in weight) metallic fuel irradiated in the Experimental Breeder Reactor was revisited using advanced electron microscopies. This fuel was irradiated to a burnup of 7.6 % at a peak cladding temperature of 521 °C. In this research, several transmission electron microscopy samples were extracted from different but representative radial locations of the fuel cross section using focused ion beam technique. Phase identification and chemical analysis with sub-micron spatial resolution was carried out using scanning transmission electron microscopy. Multiple phenomena that are critical to fuel performance, such as Zr phase and its distribution in a different matrix, were revealed in an unpreceded and highly detailed manner. An improved understanding of fuel reconstruction in U-Pu-Zr metallic fuel under reactor irradiation is provided.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Three-dimensional microstructural characterization of FBR MOX fuel and the contribution of microstructural features to the thermal conductivity of the fuel

Combination of microstructural characterization, property measurements, and phase field modeling is used to investigate fast breeder reactor (FBR) mixed oxide (MOX) fuel irradiated to burnup of 13.7% fissions per initial metal atom (FIMA). Here, the fuel was characterized at different radial locations, which revealed that grey phase can be present in the central region if it nucleates on five metal precipitates (FMPs). In addition, in the mid-radial region FMPs do not diffuse out of the region once formed and the size of Pd–Te precipitates is dictated by the porosity present in the region. Thermal conductivity measurements were conducted as a function of radial location and the microstructure of the fuel was correlated with the observed trend. Reconstructions of the 3D solid and gaseous fission product structures in different regions of the fuel were used to simulate the effective thermal conductivity (ETC) of the respective regions and determine which microstructural feature has the strongest impact on thermal conductivity. Based on conducted assessment, FMPs and Pd-Te precipitates improve local conductivity of central and mid-radial regions even in the presence of grey phase, but defects are primary contributor to the degradation of thermal conductivity on the periphery of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Manufacturing porous U10Zr fuels with controlled porosities by SPS and thermal properties

To accommodate the swelling of metallic fuels induced by the fission gas release during burnup in sodium fast reactor for a sodium-free fuel option, advanced U-10Zr fuels with controlled porosity were designed and demonstrated by spark plasma sintering. U-10Zr fuel pellets with manufactured porosities varying from 35% up to fully dense fuel pellets have been fabricated by controlling ball milling times of the starting uranium powders, sintering temperature, pressure, and duration, and the correlation among the microstructure control – porosity – sintering conditions has been established. To further mimick the pore structure in irradiated fuels, different pore formers (NaCl and NH4HCO3) have been used to control the pore size and distribution. Microstructure characterization indicates a lamellar reaction zone of U and Zr, differing from the arc-melted U-10Zr as a result of rapid consolidation of SPS at lower temperature and short durations. The thermal conductivity of U10Zr with different porosities are also measured. Here this work demonstrates the success in manufacturing new metallic fuel forms with controlled porosities and pore distribution, which can be used as model systems to investigate the thermal transfer behavior of metallic fuels in the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Ion irradiation study of lithium silicates for fusion blanket applications

Here, this study reports on the microstructural and compositional evolutions, deuterium release and lithium loss behavior in lithium silicates. Pellets of biphasic Li 4 SiO 4 and Li 2 SiO 3 were fabricated using hot pressing of powders. Sequential irradiation with Si + , He + and D + ions was performed up to 773 K to emulate one-year 6 Li burnup in 6 Li 4 SiO 4 inside the SlimCS DEMO. Crystalline Li 4 SiO 4 phase was not observed in the irradiated depth region of all the samples in this study. In spite of full amorphization in the near-surface region, Li 2 SiO 3 irradiated to a high dose at 773 K remained crystalline in the damage peak region, suggesting that Li 2 SiO 3 is more irradiation resistant to amorphization than Li 4 SiO 4 . Radiolysis may be primarily responsible for silicate decomposition and amorphization. Data from this study also show that D release from the pellet is very efficient during ion irradiation at 773 K, which is accompanied with a significant Li loss. Thermal annealing for 10 min at 773 K for the room-temperature irradiated pellet leads to a nearly complete D release without an observable Li loss. A concept of Ni coating on tritium breeder materials is discussed with supporting data to minimize Li loss without a significant impact on tritium diffusion and release.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix

The Transformational Challenge Reactor (TCR) fuel form was designed to contain large, densely packed uranium carbonitride (UCN) tristructural-isotropic (TRISO) fuel particles within a 3D printed SiC matrix, increasing the uranium density compared to conventional TRISO fuel forms and offering full geometric freedom for core design. Here, this work summarizes initial low-burnup, high-power irradiation testing of TCR fuel materials, including loose UCN TRISO particles and integral fuel compacts with ~55% TRISO particles by volume, to evaluate fission gas retention. Fission gasses were fully retained in all loose particle tests and in integral compacts irradiated at low (<250 °C) surface temperatures. Initial testing at higher (~700–750 °C) fuel surface temperatures showed fission gas release (FGR) and complete fracture of three compacts, but no FGR was observed in later high temperature tests (~300–750 °C) of both fueled compacts and loose TRISO particles. Calculated thermal stresses in the failed compacts were far less than the measured strength of the SiC matrix and the stresses in some failed compacts were less than those in compacts that did not show FGR. Thermal stress-induced matrix cracks also would not cause complete fracture because the tensile stresses transition to compression in the higher temperature regions. Therefore, fuel failure was likely not caused by thermal stresses and may have been related to leakage currents from the electrical heaters and erratic fuel surface temperatures that were only observed in the test for which failure was observed. In any case, the matrix cracks propagated through the coatings of TRISO particles located in the high-density matrix regions on the peripheries of the compacts, resulting in measurable fission gas release. The discussion focuses on the importance of understanding matrix density distributions and the particle-matrix interface properties to prevent matrix cracks from causing TRISO particle failures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transmission electron microscopy investigation of phase transformation and fuel constituent redistribution in neutron irradiated U-10wt.%Zr fuel

Uranium-10wt.%zirconium (U-10wt.%Zr) is a primary candidate for fast reactor nuclear fuels. However, there is a lack of data characterizing neutron irradiated crystallographic phases and chemistry. In the current study, the microstructural evolution of a U-10wt.%Zr fuel neutron irradiated to a burnup of 5.7 at.% was investigated by scanning transmission electron microscopy, energy dispersive spectroscopy, and selected area electron diffraction to determine the major phases, alterations in microstructure, and variations in local chemical composition at different localities of a fuel cross-section. The current study revealed that the irradiated U-10wt.%Zr fuel was comprised of various major phases, including a-U, ß-U, and d-UZr2, as well as amorphous and crystalline solid fission product (FP) precipitates within different regions of the fuel cross-section. Regions A and A/B in the center of the fuel were comprised of U-rich, U-intermediate, and U-lean localities with a-U and d-UZr 2 composing the major phases. Regions B and C in the intermediate and peripheral fuel localities, respectively, were comprised of a-U grains, U-rich (ß-U) grains with Zr-rich precipitates, U-intermediate grains (d-UZr2), and solid FP precipitates. The major phases identified were associated with the nanoscopic chemical concentrations, the phase diagram, and the as-characterized specimen temperature, with the exception of the non-equilibrium ß-U phase identified in region B. The U-lean localities in regions A, A/B, B, and C were Zr-enriched pathways along subgrain/grain boundaries. This phenomenon suggests that Zr is susceptible to radiation-induced segregation in U-Zr fuels and indicates a new mechanism for constituent redistribution.

36 MATERIALS SCIENCE↗
DOE OSTI2023

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 (gBs), 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. Furthermore, this study also revealed that the UC 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 occurs at gBs and UC/U-Mo interfaces, the fission gas inventory is always highest at those regions compared to the U-Mo grain interior. Consequently, at higher fission densities, it is likely that high burnup structure (HBS) development via grain refinement will begin at gBs and UC/U-Mo interfaces.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗