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Magnetism and finite-temperature effects in UZr 2 : A density functional theory analysis

The structure and the thermophysical properties of d-UZr 2 are investigated using 0 K density functional theory and ab initio molecular dynamics (AIMD). Modeling the true paramagnetic state of this intermetallic compound has been challenging using first-principles calculations. For the first time, we find that the generalized gradient approximation method without applying an on-site Coulomb interaction term (Hubbard U) can result in a ground state that is antiferromagnetic (AFM). We believe that this weak AFM ground state is the closest to the real paramagnetic state. We found that structure optimization at finite temperatures using AIMD is necessary to achieve this ground state instead of the non-magnetic and ferromagnetic states previously reported in the literature. Our findings indicate that applying the Hubbard U on uranium f-orbitals in this metallic system is unnecessary and not recommended, as it leads to a large overestimation of the volume and introduces an unphysical strong spin polarization. Finally, our approach results in atomic volume, thermal expansion, and heat capacities that have strong agreement with experiments.

36 MATERIALS SCIENCE↗

Mesoscale modeling of microstructure-dependent thermal conductivity in U-Zr fuels

In uranium-zirconium (U-Zr) based metallic fuels, different phases can form at different compositions and temperatures. Typically, lamellar δ-UZr 2 and α-U phases are the dominant microstructures in U-rich U-Zr alloys at temperatures below 880 K. In this work, a finite element method based mesoscale modeling technique is used to calculate the effective thermal conductivities of such heterogeneous microstructures, using the thermal conductivities of two individual phases and their interphase thermal resistance (Kapitza resistance) as input parameters. The Kapitza resistance between δ-UZr 2 and α-U is determined at different temperatures, which shows an approximately T 3 dependence in the temperature range between 300 and 800 K. In addition, the Kapitza resistance exhibits a strong dependence on the aspect ratio of the δ-UZr 2 phase. Further, an analytical model is therefore developed to quantify the effects of both temperature and δ-UZr 2 aspect ratio on the Kapitza resistance. Using this newly developed Kapitza resistance model, the effective thermal conductivities of a number of δ-UZr 2 + α-U heterogeneous microstructures in U-Zr alloys, including non-lamellar microstructures, can be estimated accurately.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Thermal Calculations to Determine Maximum Zircaloy Temperatures During Reactor Operations for Campaign 1 Fuel Assemblies

The H-Canyon Facility plans on processing bundles from Non-Aluminum Spent Nuclear Fuel (NASNF) Risk Group 1 (UO 2 core fuels). Uranium-zirconium (UZr) alloys and zirconium clad UO 2 fuel containing UZr intermetallic compound have the potential to cause explosions during dissolution. UZr intermetallic typically only forms when uranium and zirconium (or zirconium alloy) cladding are in contact at high temperatures. Thermal modeling was performed to determine the maximum cladding temperature for Campaign 1 Zircoloy cladded fuel assembles during reactor operation to help determine if the UZr intermetallic compound could have formed during reactor operations. The temperature of 350°C was chosen as a conservative minimum value for the possible formation of intermetallic UZr compound. The modeling approach, inputs, and results are outlined in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Irradiation Effects on Stability of δ-UZr2 phase in U-50 wt% Zr Alloy

U-50wt%Zr is a candidate metallic nuclear fuel with potential application in light water reactors due to its excellent thermal properties and high radiation tolerance. The Zr-rich UZr fuels possess greater swelling resistance and fission gas release characteristics compared with U-rich UZr fuels. In this current study, the δ-phase U-50wt%Zr is proton irradiated at room temperature to 1 displacement per atom (dpa) to provide insights on phase stability under irradiation conditions. High resolution characterization of Transmission Electron Microscopy (TEM) and Atom Probe Tomography (APT) characterization techniques are used to elucidate microstructural changes due to irradiation. TEM and APT results show highly oriented bcc β-Zr-rich platelet precipitates nucleating adjacent to α-U phases inside the UZr 2 matrix. Formation of this platelet morphology is characteristic of Widmänstatten structure which can be attributed to a variety of factors such as differences in thermal expansion coefficient between the phases, grain size, alloy composition, and cooling rate. The phases present are distinctly different than those observed through in situ annealing, but irradiation accelerates diffusion and phase separation kinetics. These microstructural changes in U-50wt%Zr are different from those achieved by pure thermodynamic or high temperature heavy ion irradiation experiments. Finally, our work, together with previous ones, highlight the necessity to study the U-Zr phase diagram under non-equilibrium thermodynamics conditions to support this material's deployment as a viable nuclear fuel form.

36 MATERIALS SCIENCE↗

Investigation of γ - ( U , Z r ) structural properties and its interfacial properties with liquid sodium using ab initio molecular dynamics

In this study, the elastic properties, structural parameters, sound velocity, and Debye temperature of γ–(U,Zr) were computed using ab initio molecular dynamics (AIMD) at temperatures between 1000 K and 1400 K and for Zr content between 0 at. % and 100 at. %. UZr is used as a metallic fuel for Sodium Fast Reactors (SFRs). The study of the mechanical and thermal behavior of these alloys leads to a better data-informed fuel design. The bulk modulus, shear modulus, Young's modulus, and Poisson's ratio were calculated from the elastic constants and their dependence on Zr content and temperature was investigated, comparing the results with previous computational work and the available experimental data in the literature. Interfacial properties between UZr (up to 32 at. % which typically exists in nuclear fuel) and liquid sodium are also of interest due to the presence of a sodium bond between the fuel and the cladding in metallic nuclear fuel. The interfacial energy between γ–(U,Zr) and liquid sodium, the surface tension of liquid sodium, and the work of adhesion were computed at different temperatures and Zr concentrations. It was demonstrated that γ–(U,Zr) is completely wetted by liquid sodium at all the investigated temperatures and Zr concentrations. Finally, this work provides the basis for the determination of interfacial resistances in SFRs and their implementation into heat transfer fuel performance simulations, which will be the subject of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Diffusion study of uranium mononitride/zirconium carbide composite for space nuclear propulsion

Here, the next generation of space exploration will require extensive developments in rocket technology. Space nuclear propulsion is of interest due to its high fuel density and power, however, it also has high temperature and stability demands. This study examines a uranium nitride/zirconium carbide ceramic-ceramic particle composite as a potential fuel for these missions. Based on the constituent properties, this fuel composite is expected to be highly thermally efficient and resistant to the hot hydrogen propellant. One of the main concerns with this composite is unknown diffusion behavior between UN and ZrC over time. Diffusion couples and composites were used to determine how these constituents will behave when left in contact at high temperatures. UN and ZrC were found to have limited but observable diffusion at the phase boundary. The resulting phase is a UZr(CN) quaternary phase. UZr(CN) has been examined for other high temperature and gas nuclear reactors with preliminary success; however, there lacks sufficient data to fully qualify this phase. Based on the limited information available, the resultant quaternary phase could be deemed acceptable and even provide better adhesion for the fuel particles (UN) to the matrix (ZrC).

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Casting and Characterization of U-50Zr

Uranium alloyed with 50 wt% zirconium (U-50Zr) is a proposed light water reactor (LWR) nuclear fuel by Lightbridge Corporation (LTBR). The proposed method for making the U-50Zr alloy is to arc-melt master alloys and then remelt in a vacuum induction melter (VIM) to consolidate the material and cast into an intermediate shape. After casting, the material will eventually need to be formed into a desired fuel shape. The work discussed in this report resulted from a joint effort between Ltbr and Pacific Northwest National Laboratory (PNN) to investigate a 500g – 1kg scale casting process to produce the U-50Zr alloy in the desired δ-UZr 2 phase and characterize the impurities and microstructure that result from the casting process. Master alloys were fabricated in an arc melter, then five castings were carried out in a VIM with multiple inert coating and crucible materials to find an appropriate combination. Both ZrO 2 and graphite crucibles were used and different combinations of Y 2 O 3 , CaZrO 3 , and TiC to identify which would contain the molten metal with the least reaction. On each casting, the C, O, N, H impurities were analyzed as well as the phase by x-ray diffraction and microstructure. Of the five castings, two resulted in majority of δ-UZr 2 phase-pure material and had impurity levels within acceptable ranges. The two most successful castings utilized a graphite crucible with a TiC undercoating and a Y 2 O 3 overcoat. The O and N levels were below 1000 ppm and the C content was variable but did not result in measurable carbide formation. The highest success casting resulted in an average of 282 ppm C, 567 ppm O, 217 ppm N and 79 ppm H. This casting's crucible and inert coating material was repeated with slightly different casting parameters and resulted in higher C numbers but similar phase identification. The differences between each casting are discussed and recommendations are made for future experiments to better decide on a casting process to go forward with.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Phase-field simulations to inform nuclear fuel performance modeling

Software tools to simulate nuclear fuel performance at the engineering scale, such as Idaho National Laboratory (INL)’s BISON code, are increasingly relied upon in regulatory and economic decision-making. However, accurate results from these tools depends on the availability of materials parameters that are used as input. In recent years, atomistic and mesoscale simulation methods have emerged as a cost-effective, expedient means to obtain such input parameters. Phase-field simulations using INL’s Marmot application have been used to obtain microstructure-level parameters and to improve material models for fuel performance modeling using BISON. In this talk, recent examples of this process are given, including applications in UO2, U3Si2, and UZr fuel.

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Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding

As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. In this work, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr 2 phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal ( U ,Zr) 6 Fe phase (space group I4/mcm) and cubic ( U ,Zr)( Fe ,Cr) 2 phase (space group Fd 3 ¯ m). The ( U ,Zr)( Fe ,Cr) 2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm 3 ¯ m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4 2 /mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.

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Phase evolution of U-Zr system in a thermal cycling neutron diffraction experiment: as-cast U-35Zr and U-50Zr

Here, phase evolution of as-cast U-35 wt% Zr and U-50 wt% Zr alloys during thermal cycling (303–1073 K) was investigated using in-situ neutron diffraction. Analysis was performed using Rietveld crystal and microstructure refinements from time-of-flight neutron diffraction data, with a focus on evolution of lattice parameter, atom ordering, unit cell volume, and weight fractions during the thermal cycling. Disordered δ-UZr 2 and residual γ-(U,Zr) phase are retained in the field of δ of the U-Zr equilibrium phase diagram for the U-50Zr sample, and in the field between δ and α-U for the U-35Zr sample. The α phase that is present in the reported phase diagrams of U-Zr was not observed in the diffraction patterns. The evolutions of lattice parameter and unit cell volume of δ and γ are affected by both thermal expansion and chemistry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Impacts of annealing treatment on the microstructure of U-Mo monolithic fuel plates

The objective of this paper is to analyze the effect of different heat treatments on the fuel microstructure and texture in laboratory scale monolithic U-Mo fuel plates. Such analyses are relevant as they could possibly inform future fabrication methods for the optimization of microstructure and of material properties to ultimately improve fuel performance during irradiation. For this reason, detail characterization including Energy x-ray dispersive spectroscopy (EDS) and electron backscattered diffraction (EBSD) techniques were applied on U-Mo plates fabricated with different heat treatments to understand the influence of fabrication on fuel microstructure. Here, U-Mo fresh fuel specimens with or without annealing treatments (homogenization and stress release annealing) after rolling were analyzed in this work. In this study for the first time EBSD was systematically used to analyze the heat treatments influence on texture present in fresh fuel monolithic U-Mo fuel plates. Such changes were connected to other known microstructural and chemical changes. In this work minimization of molybdenum concentration variation and gamma phase decomposition of the U-Mo fuel core after homogenization was observed. Also, an increase in the interaction layer of the U-Mo fuel core with the Zr interlayer diffusion barrier was observed with the UZr 2 growing up to 1 µm in thickness. Such microstructural changes were aligned to the changes in texture and grain structure. Indeed, common texture observed in body-centered cubic (BCC) metal after cold rolling (e.g., α, ξ, γ fibers) were minimized by the annealing/homogenization process. Moreover, grain structure was influenced by the fabrication route with elongated grains formed in the rolled sample and equiaxed grains found after annealing. Such features are relevant since the formation of microstructural features formed during fuel fabrication can impact fuel performance in reactor.

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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↗

Manufacturing porous U-10Zr metallic fuels with controllable microstructure by volume control spark plasma sintering

In this paper, the volume control spark plasma sintering tool has been designed and applied to sinter porous U-10Zr metallic fuels, by which the sintered sample volume can be precisely controlled. Ethanol and NH 4 HCO 3 are used to control the powder compact or as pore formers to control the pore size and pore structure. Without pore formers, the fuel pellet displays an inhomogeneous microstructure consisting of highly porous and highly densified areas. Uneven powder stacking in the green body results in a non-uniform microstructure, and in the closed-packed area, Joule heating accelerates the neck formation and densification. The addition of ethanol reduces the friction between the powders, resulting in isolated pores formed by the stacking of powders during the sintering. By adding NH 4 HCO 3 , the pore size, and structure can be well controlled, and an interconnected pore structure can be obtained upon the decomposition of the NH 4 HCO 3 . Further, a uniform microstructure and pore distributions can be achieved through the U-10Zr fuel pellets by controlling current flow during the volume control SPS sintering. The microstructure and phase characterization of the sintered porous U-10Zr pellets show major phases of α-U and α-Zr for the sample with short dwelling. For the sample with long dwelling (30 min), the ω UZr 2 in the Zr-enriched area has been observed. The strategy of volume control SPS sintering with the assistance of pore formers could be used to fabricate porous U-10Zr metallic fuels to mimic the microstructure evolution of irradiated metallic fuels (including porosity) and could enable a possible solution for the design of new sodium-free metallic fuels for high burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Annealing influence on the microstructure of irradiated U-Mo monolithic fuel foils

In this study we compared the microstructure evolution of U-Mo fuel foils produced with and without heat treatment at low burn-up via advanced post-irradiation examination. The aim of this study is to observe after irradiation the ways in which the fabrication processes have influenced fuel behavior at early-stage irradiation, as for very low burn up microstructural studies are lacking. In this work it was observed that the larger grain size detected in the heat-treated samples before irradiation led to decreased grain refinement after irradiation. Grain refinement was associated with the presence of small nano-size bubbles and precipitates. This phenomenon is hypothesized to influence early fuel swelling during reactor irradiation. Grain refinement was also observed to increase in regions where γ-U decomposition was present. Thus, it was enhanced in the samples fabricated without heat treatment. The heat treatment also increased the thickness of the U-Mo/Zr interface, namely of the UZr 2 layer. However, the influence of this layer on fuel performance needs further investigation. On one side, it may contribute to better mechanical bonding; on the other, it may influence swelling and blistering in the interaction layer as porosity increases when this layer is increased. This was observed especially in the presence of increased area containing low Mo concentration, and thus containing a higher fraction of the α-U phase, which is highly susceptible to irradiation induced swelling. Strong evidence of reverse transformation under irradiation (α-U + γ'-U2Mo → bcc γ-U) was observed in these samples. While the precipitates (carbides and oxide) seem to be unaffected by the irradiation at these low burnups. However, further analyses are necessary at higher burn-up to assess the exact impact different heat treatments have on fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Impacts of irradiation-induced nanostructure on phonon linewidths and thermal conductivity in U-Zr alloy

Ion irradiation can significantly change material properties. Thermal conductivity degradation of reactor fuel is detrimental to safety and efficiency—driving research on how irradiation impacts heat-carrying phonons. Here, we report the measurements of phonon dispersion and lifetimes (inverse linewidths) in freestanding irradiated crystals and reveal the relationship between thermal conductivity and proton irradiation induced nanostructures in UZr 2.6 using inelastic x-ray scattering, electron microscopy, modulated thermoreflectance, and first-principles calculations. Irradiation results in more significant increases in phonon linewidths and decreased thermal conductivity in the basal plane than that along the c axis, decreasing thermal conductivity anisotropy. Furthermore, microscopy reveals that this results from irradiation induced nanoprecipitates and metastable phase separation. Surprisingly, irradiation leads to smaller phonon linewidths near some zone boundaries, which we attribute to irradiation induced short-range order.

36 MATERIALS SCIENCE↗

A physics-based model of cladding wastage layer formation rate

In this report, lower length scale simulations to inform an engineering-scale model of fuel-cladding chemical interaction (FCCI) conducted under the auspices of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in FY22 are described. An overall strategy for the implementation of the BISON model is described, and the past and current lower length scale work on FCCI formation is put into the context of this overall strategy. Density functional theory and kinetic Monte Carlo simulations are used to determine the diffusion coefficient of neodymium in iron. Density functional theory calculations are also used to parameterize a parabolic approximation for the dependence of free energy on composition for the intermetallic compound Fe 17 Nd 2 . A phase-field model of the Fe-Nd system was developed that includes the random solid solution body-centered cubic phase and the intermetallic Fe 17 Nd 2 . The model was used to simulate growth of the intermetallic layer in a diffusion couple and the results were compared to experiment. The model’s prediction of the parabolic growth constant is within 40 % of the experimental value. The model is also used to simulate the formation of a grain boundary attack layer that is observed in experiment. The simulations show enhanced Nd concentration along the grain boundaries, but do not clearly show formation of a Fe 17 Nd 2 layer

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗