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DOE OSTI2023

Development of Mechanistic Fission Gas Release and Swelling Models for UN Fuels in BISON

This report describes the work in NEAMS (Nuclear Energy Advanced Modeling and Simulation) to develop a mechanistic fission gas model for uranium nitride fuels in BISON. The existing Sifgrs (Simple integrated fission gas release and swelling) model tracks the average properties of two bubble populations in the bulk and at the grain boundaries. It was recognized that dislocations play a crucial role in the fission gas swelling of UN, and an irradiation-induced dislocation density model was needed, as well as a model describing how fission gas interacts with dislocations creating a third population of bubbles along dislocations. A mechanistic model that tracks dislocation bubbles was implemented in Sifgrs. This model was used to simulate fission gas swelling and release in UN. Lower-length-scale calculations and experimental observations from carbide fuel were leveraged to help populate the model with essential parameters. As a placeholder, an empirical function was formulated for the evolution of the dislocation network. Because of the difference in evolution of this network at different temperatures, the dislocation bubbles are able to capture behavior that the bulk intragranular bubbles cannot. It was found that the bulk bubbles dominate microscopic swelling at low temperatures, and the dislocation bubbles dominate at higher temperatures. Based on this model, the transition between the two bubble types is the main factor behind the breakaway swelling phenomenon in UN. In order to test the model, a couple of assessments were run. For the lower temperature JOYO pins, the model produces reasonable fission gas release and swelling values. For the higher temperature SP1 pin, the model dramatically underestimates the fission gas release. This issue is attributed to the gas being trapped inside dislocation bubbles, unable to escape to grain boundaries to cause fission gas release, and could be remedied by a mechanistic dislocation model that allows the dislocation density to decrease at very high temperatures. A preliminary mechanistic dislocation model was developed supported by first-principle calculations. These calculations provided valuable insight into how interstitial defects cluster in UN in the {110} orientation, and may eventually form dislocation loops, leading to the conclusion that dislocation loops may nucleate from these clusters. An estimate of the dislocation line energy in UN was obtained and will be improved in future work. The free-energy cluster dynamics code Centipede was used to track interstitial and vacancy absorption at dislocation loops and calculate their growth. In addition, improvements to Centipede were made including new convergence criteria for transient simulations and sink driving force updates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Redistribution of radionuclides in irradiated AGR-1 UCO TRISO fuel after 1800 °C safety testing

Release of radionuclides from intact tristructural-isotropic (TRISO) coated particle fuel at normal and accident conditions is a primary metric of fuel performance. The distribution of fission products and actinides in the TRISO layers of individual particles provides insight on radionuclide transport and release behavior and was determined using scanning electron microscopy analysis. Particles were isolated from an irradiated fuel compact (AGR-1 Compact 4-4-2) and analyzed as-irradiated or after individual particle safety-testing at 1800 °C for 650 h. Particles were selected for comparison based on their remaining 110mAg fission product inventory. These comparisons corroborated the observation that the 110m Ag inventory is a marker for relative irradiation temperature based on observed radionuclide distribution in the SiC layer. The comparison also indicated that the in-pile behavior influences the fission product and actinide species interactions with the TRISO layers during high temperature exposure after irradiation. The analysis confirms both palladium and uranium diffusion, as well as other species, are active in the UCO TRISO fuel system at 1800 °C and that palladium transport is active at lower temperatures relative to uranium. While diffusion across the SiC layer was observed, the intact nature of the SiC layer after the 1800 °C, 650-h exposure indicates the SiC layer maintained its functionality as a fission product barrier by mitigating release of radionuclides at beyond accident margin temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Characterization Report for the MP-1 Experiment Fabrication Campaign (Rev.1)

This report covers the results of MP 1 fuel characterization analysis to provide a detailed understanding of the as-fabricated fuel that would be irradiated in the MP 1 (Mini-Plate 1) experiment. The work was performed at Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and Los Alamos National Laboratory (LANL). Characterization of the as-fabricated fuel was performed in accordance with Characterization Plan for the Fabrication of U 10Mo for the MP 1 Experiment, hereinafter the MP 1 Characterization Plan (MAQP 2016). Under the MP 1 Characterization Plan, fuel foils at different points in the foil fabrication process (e.g., hot rolled annealed, cold rolled annealed) were studied to understand the effect of various thermomechanical processes on the fuel microstructure. Samples were sent to each of the three organizations so that testing and analysis could be performed independently using similar equipment and a standardized set of measurement and analysis procedures. The MP 1 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare independently obtained results, and achieve a three-way validation. PNNL characterized 28 uranium–10 wt% molybdenum (U 10Mo) samples. Six pieces/specimens from each sample/foil were sectioned, in accordance with the MP 1 Characterization Plan. Similarly, INL and LANL also examined 24 and 17 U 10Mo specimens, respectively. These total 69 samples consist of six types of MP 1 characterization foils fabricated by Babcock and Wilcox Technologies (BWXT) from ingots produced by the Y-12 National Security Complex: (a.) 0.047 in.-thick hot-rolled and annealed samples with and without Zr layers; (b.) 0.02 in.-thick cold-rolled and annealed samples with and without Zr layers; and (c.) 0.01 in.-thick cold-rolled and annealed samples with and without Zr layers. Microstructure, chemical composition, carbide morphology, U 10Mo foil thickness, Zr thickness, mechanical properties (microhardness), and density were evaluated in both longitudinal and transverse directions for foils of the three different thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Intentional Uranium Tagging for Material Provenance and Pathway Forensics (LA19-Intentional-Forensics-NDD3Bb) (Annual Report for FY21)

This report describes the outcomes of the third and final year of a project to research the feasibility of tagging uranium materials, especially nuclear fuels. The experimental focus remained on metallic uranium forms under prospective surface and bulk tagging scenarios. Overall, the results showed that the tags could be successfully imparted and characterized in both as-built and degraded conditions. This was in line with expectations coming into this project, indicating promise for both surface and bulk tagging of metallic forms of uranium. Multiple surface tagging techniques and detection strategies were explored in FY21, with an emphasis on improving tag quality, readability, and detection in the field. Non-radioactive materials were used as a testbed. Selective deposition via laser beam was determined to be successful in imparting a readable titanium deposit on a stainless steel base plate, and can be read with high resolution characterization techniques (e.g., scanning electron microscopy) and field capable tools (e.g., eddy current testing). Other deposition techniques, such as selective deposition via electron beam and photoluminescent tags, were explored in FY21, and while success for these techniques would be dependent on additional work, these techniques showed potential for surface tagging applications. To survey bulk taggant elements for bulk uranium metal, 16 tagging elements were spread among 18 depleted uranium castings (4 baseline, 3 mix, 1 dilution, and 10 recycle). Most of these were made and characterized in FY21. Taggant acceptability was based upon manufacturability, detectability, and persistence from the standpoint of two detection options: bulk chemical analysis (for “chemical taggants”) and microstructural analysis (for “second phase taggants”). Taggant detection in both up-front manufacturing and in the face of “degradations” such as dilution, mixing, and recycling was generally good. Two independent laboratories carried out chemical analysis on most of the castings, and often at several locations within a casting, and the results are discussed. Scanning electron microscopy+EDS microanalysis revealed the second phases mostly contained the expected tagging elements. The shapes and 2 spatial distributions of these micron-sized carbides, oxides, and intermetallic particles offers opportunities for further science-based investigation and tagging optimization. Overall, V and Co currently appear as the best choices for chemical taggants while Al, Ti, Mn, Co, Pd, and Tb all look good as second phase taggants. The other elements considered here – Sc, Ni, Ge, Nb, Ce, Ta, W, Ir, and Au – while not being ruled out, require more study to become viable options. It is of special note that in the recycling study only one of the 12 elements fell out of detection even after 10 meltings, demonstrating their persistence. An Appendix tabulates all chemical analysis results to enable more quantitative and statistical studies of detection opportunities and limitations, as a part of a related project.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Microstructural-Level Fuel Performance Modeling of U-Mo Monolithic Fuel

As the physics that governs the microstructural evolution of nuclear fuel span various time and spatial scales, to fully understand the fuel behavior inevitably involves atomic to mesoscale resolution that can be difficult to determine experimentally. Microstructural-level modeling and simulations can be used to develop physics-based materials models that can provide physical understanding to inform fabrication process control, as well as a valuable feedback mechanism between post-irradiation examination (PIE) results and fabrication parameters. In accordance with the program schedule, the primary goals of the microstructure modeling effort are to: 1. Address critical microstructural questions and provide practical guidance to the fabricator via the fuel product specification; 2. Provide mechanistic inputs for the existing fuel performance code to improve its descriptive and predictive capability at the macroscopic scale. In fiscal year (FY)-21, the work scope consisted of six main facets: (1) the effect of carbides on fuel performance; (2) gas diffusivity in different phases; (3) integration of microstructural fuel performance modeling; (4) property degradation; (5) irradiation creep; and (6) historical analysis of microstructure data. Brief summaries of each are included below.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

AGR-5/6/7 Irradiation As Run Predictions Using PARFUME

The PARticle FUel ModEl (PARFUME), a fuel performance modeling code used for high-temperature gas-cooled reactors, was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using as-run physics and thermal data. The AGR-5/6/7 irradiation test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter northeast flux trap position of the Advanced Test Reactor (ATR) at Idaho National Laboratory. Each capsule contains compacts filled with uranium oxycarbide unaltered fuel particles. This report documents the calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated fission product release fraction from the fuel. The calculations include modeling of the AGR 5/6/7 irradiation that occurred from February 2018 to July 2020 over nine ATR cycles, including six normal cycles and three power axial locator mechanism cycles, for a total of approximately 376 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 3, and 4. Fuel particle failures were predicted in two of the compacts in Capsule 2 and one particle failure is predicted in each one of the compacts in Capsule 5. All compacts that exhibited fuel particle failures predicted by PARFUME were caused by localized stress concentrations in the silicon carbide (SiC) layer caused by cracking in the inner pyrolytic (IPyC) layer. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. Compacts with a lower irradiation temperature and fluence experienced the smallest buffer-IPyC gap formation. Conversely, higher irradiated temperature compacts with a high fluence experienced the largest buffer-IPyC gap formation. Compact 3-6-3 experienced the largest buffer IPyC gap formation of just under 21.7 µm. The release fraction of fission products silver (Ag), cesium (Cs), and strontium (Sr) vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 3, reaching up to 59.5% for the TRISO fuel particles (compact 3-6-3). The release fraction of the other two fission products, Cs and Sr, are much smaller. A maximum Cs release fraction of 1.1% occurred in compact 3-4-3 and 4.4% for Sr in compact 3-6-3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

USc 2 C 2 and USc 2 NC Clusters with U–C Triple Bond Character Stabilized Inside Fullerene Cages

The chemistry of f-block metal–carbon multiple bonds is underdeveloped compared to well-established carbene complexes of the d-block transition metals. Herein, we report two new actinide-rare earth mixed metal carbides and nitrogen carbide cluster fullerenes, USc 2 C 2 @D 5h (6)-C 80 and USc 2 NC@D 5h (6)-C 80 , which contain U–C bonds with triple bond character and were successfully synthesized and characterized by mass spectrometry, UV–vis–NIR spectroscopy, Fourier transform infrared spectroscopy, single crystal X-ray diffraction, and DFT calculations. Crystallographic studies show that the two previously unreported clusters, USc 2 C 2 and USc 2 NC, are stabilized in the D 5h (6)-C 80 carbon cage and adopt unique trifoliate configurations, in which C 2 /NC units are almost vertically inserted into the plane defined by the U and two Sc atoms. Combined experimental and theoretical studies further reveal the bonding structure of USc 2 C 2 and USc 2 NC, which contain C=U(VI)=C and C=U(V)=N bonding motifs. The electronic structures of the two compounds are determined as U 6+ (Sc 2 ) 6+ (C 4– ) 2 @D 5h (6)-C 80 4– and U 5+ (Sc 2 ) 6+ (N) 3– (C) 4– @D 5h (6)-C 80 4– , respectively. Quantum-chemical studies confirm that the U–C bonds in both molecules show unprecedented multicenter triple-bond character. Furthermore, the discovery of this unique U–C multiple bond offers a deeper understanding of the fundamentals of uranium chemistry.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Alternate Methods for Cleaning Zirconium Plate

The United States High Performance Research Reactor Conversion Program identified the need to transition from highly enriched uranium fuel to lowly enriched uranium fuel for high power research and civil reactors. One of the fuel configurations selected for these reactors was a uranium (U) foil alloyed with 10% molybdenum (Mo). The U-10Mo fuel foil was then covered with a zirconium (Zr) interlayer and pressed into an aluminum cladding. Idaho National Laboratory prescribed a need to remove surface oxidation on the Zr prior to bonding it with U-10Mo fuel foil. To determine the best method of oxidation removal, a variety of polishing methods, including hand polishing with a diamond paste or Scotch-Brite™ pads, mechanically wet-polishing with abrasive belts, and chemical etching with hydrofluoric/nitric acid mixtures, were investigated on pre-polished Zircaloy 702 sheets purchased directly from the manufacturer. No change in surface roughness was detected using any of these methods relative to the as-received material. In fact, scanning electron microscopy detected grit from hand polishing with Amplex Grade 30 waste soluble diamond paste and Scotch-Brite™ pads. Visible staining was found, and fluorine detected via X-ray photoelectron spectroscopy on the flash etching samples. These contaminants were not seen on the as-received material. In addition to the need to remove the oxide layer on the Zr prior to co-rolling, there are concerns that the Neolube debonding agent on the co-rolling can may contaminate the Zr surface after co-rolling. The experimental results found that zirconium in contact with Neolube coated steel plates was easily cleaning with ethanol wiping after heat treated and no carbide inclusions remained on the surface. Overall, all cleaning methods tested did not significantly change the surface roughness of the plates or reduce the oxide layer present on the zirconium. Carbon was found on all the samples, before and after cleaning, except for the flash etch samples which detected fluorine instead. The insignificant changes seen in the cleaned plates relative to the as received plates question the need for cleaning prior to rolling. Further investigation in the requirements for cleaning at this step is recommended. Current methods for cleaning the surface after hot rolling with alcohol appear sufficient. After hot rolling, if Neolube does peel off the can and onto the zirconium coated foil surface, a dry abrasive or diamond paste polish is not recommended due to the chance of abrasive being embedded into the surface. However, the ability of an etchant to lift the lubricant off the zirconium surface indicates that a spot etch method could work to clean specific spots where Neolube is present, if the need arises.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

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↗