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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Uranium Single Particle Analysis for Simultaneous Fluorine and Uranium Isotopic Determinations via Laser-Induced Breakdown Spectroscopy/Laser Ablation–Multicollector–Inductively Coupled Plasma–Mass Spectrometry

Uranyl fluoride (UO 2 F 2 ) particles (<20 μm) were subjected to first-of-its-kind analysis via simultaneous laser-induced breakdown spectroscopy (LIBS) and laser ablation multi-collector inductively coupled plasma–mass spectrometry (LA–MC–ICP–MS). Briefly, a nanosecond pulsed high-energy laser was focused onto the sample (particle) surface. In a single laser pulse, the UO 2 F 2 particle was excited/ionized within the microplasma volume, and the emission of light was collected via fiber optics such that emission spectroscopy could be employed for the detection of uranium (U) and fluorine (F). The ablated particle was simultaneously transported into the MC–ICP–MS for high precision isotopic (i.e., 234 U, 235 U, and 238 U) analysis. This method, LIBS/LA–MC–ICP–MS was optimized and employed to rapidly measure 80+ UO 2 F 2 particles, which were subjected to different calcination processes, which results in varying degrees of F loss from the individual particles. In measuring the particles, the average F/U ratios for the populations treated at 100 and 500 °C were 2.78 ± 1.28 and 1.01 ± 0.50, respectively, confirming loss of F through the calcination process. The average 235 U/ 238 U on the particle populations for the 100 and 500 °C were 0.007262 (22) and 0.007231 (23), which was determined to be <0.2% from the expected value. The 234 U/ 238 U ratios on the same particles were 0.000053 (11) and 0.000050 (10) for the 100 and 500 °C, respectively, <10% from the expected value. Notably, each population was analyzed in under 5 min, demonstrating the truly rapid analysis technique presented here in this paper.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Transient behavior of oxide fuels with controlled microstructure and Cr2O3 additive

Abstract Microstructure and Cr 2 O 3 doping profoundly impact the thermal-mechanical properties and fracture of oxides fuels. It is a challenge to study the transient behavior of nuclear fuels under loss-of-coolant-event (LOCA). In this study, the crack behavior of UO 2 pellets with controlled grain structure and Cr 2 O 3 doping was tested with rapid power ramping (300−900 °C per min) mimicking a prototypical LOCA heating profile. Dense micron-sized UO 2 pellets display well-maintained integrity without cracking with the ramping up to 1500 °C at a heating rate of 8 °C per second. Fracture occurs in both pure and Cr 2 O 3 -doped dense nano-sized UO 2 pellets. The Cr 2 O 3 doped oxide fuel pellet with a larger grain size (~ 22.2 μm) displays the best performance under LOCA testing due to its highest thermal conductivity under high temperature. FEA calculations suggest a temperature gradient across the fuel pellet during transient testing, resulting in residual stress and cracking, which can be correlated with their thermal-mechanical properties.

Zhao, Dong (ORCID:0000000191377257)↗
DOE OSTI2024

The quest for safer nuclear fuels

Here, to increase the safety and performance of nuclear reactors, researchers are working on developing ‘accident tolerant’ fuels. Ideally, these fuel pellets will be less susceptible to swelling and be more resistant to extreme heat. This led chemists to develop uranium dioxide (UO 2 ) doped with divalent elements, E 2+ . These dopants can introduce vacancies into the lattice that lead to larger grain sizes, which create longer pathways for fission gasses to migrate out of the grains. In undoped fuel, fission gas build-up at grain boundaries causes the fuel pellet to swell. Therefore, having larger grains that slow fission gas migration into grain boundaries is highly desired. Divalent Cr 2+ ions are a commonly explored dopant for UO 2 fuels. These materials have larger grain sizes while maintaining the dissolution kinetics of undoped UO 2 , making their long-term storage and disposal paths similarly safe. Atomistic models have suggested that Mn 2+ is a conceivable alternative doping material, with the potential to increase the sizes of grains even more than Cr 2+ dopants. As such, experimental studies are required to understand the physical chemistry and reactivity of these potential fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Chemical bonding in americium oxides probed by X-ray spectroscopy

The electronic structure and the chemical state in Am binary oxides and Am-doped UO 2 were studied by means of X-ray absorption spectroscopy at shallow Am core (4d and 5d) edges. In particular, the Am 5f states were probed and the nature of their bonding to the oxygen states was analyzed. The interpretation of the experimental data was supported by the Anderson impurity model (AIM) calculations which took into account the full multiplet structure due to the interaction between 5f electrons as well as the interaction with the core hole. The sensitivity of the branching ratio of the Am 4d 3/2 and 4d 5/2 X-ray absorption lines to the chemical state of Am was shown using Am binary oxides as reference systems. The observed ratio for Am-doped UO 2 suggests that at least at low Am concentrations, americium is in the Am(III) state in the UO 2 lattice. To confirm the validity of the applied AIM approach, the analysis of the Am 4f X-ray photoelectron spectra of AmO 2 and Am 2 O 3 was also performed which revealed a good agreement between experiment and calculations. As a whole, AmO 2 can be classified as the charge-transfer compound with the 5f occupancy (n ƒ ) equal to 5.73 electrons, while Am 2 O 3 is rather a Mott–Hubbard system with n ƒ = 6.05.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Single crystal structure and photocatalytic behavior of grafted uranyl on the Zr-node of a pyrene-based metal–organic framework

Accurately characterizing actinide oxides bound to metal–organic frameworks (MOFs) is important for designing MOFs as radioactive waste sorbents and catalytic supports. Here, the zirconium MOF NU-1000 was post-synthetically modified through solvothermal deposition to include the uranyl (UO 2 2+ ) ion and characterized via single-crystal X-ray diffraction. Additonally, bond lengths derived from the diffraction pattern and Raman spectroscopy indicate that UO 2 2+ maintains its chemical properties upon deposition, while alcohol oxidation photocatalysis reveals photo-interactions between the pyrene linkers and the UO 2 2+ ion.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Electrochemical behaviour of uranium at a tripolyphosphate modified ITO electrode

UO 2 2+ binds to the surface of a tripolyphosphate modified mesoporous indium tin-doped oxide electrode (nanoITO|P 3 ). Electrochemical studies reveal that nITO|P 3 electrodes catalyze the 2-electron interconversion between UO 2 2+ and U 4+ with the P 3 -ligand assisting in the rate-limiting proton-coupled reduction of U(V) to U(IV), based on the kinetic isotope effect (1.8). Product composition between nITO|P 3 (U 4+ ) and surface adsorbed UO 2 can be controlled by adjusting the proton concentration and/or scan rate in voltammograms. Furthermore these studies with uranium suggest that nITO|P 3 electrodes are good candidates for redox transformations with other actinides including neptunium, plutonium, and americium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Flux-assisted polytypism in the [Na 2 Cl]GaQ 2 heterolayered salt-inclusion chalcogenide family

Two polytypic heterolayered salt-inclusion chalcogenides, o-[Na 2 Cl]GaQ 2 and t-[Na 2 Cl]GaQ 2 , were obtained via a NaCl/NaI flux-assisted synthesis, as part of an investigation of the Na–Ga–Q (Q = S and Se) system. The use of a different flux, NaBr/NaI, in the Na–Ga–Se system did not lead to the formation of salt-inclusion phases, but instead the novel Na 2 GaSe 3 and Na 4 Ga 2 Se 5 phases were obtained. Thermal and electronic properties of [Na 2 Cl]GaS 2 materials were investigated with differential scanning calorimetry, post-quenching ex situ powder X-ray diffraction (PXRD), high-temperature PXRD, and enthalpy and electronic structure calculations via density functional theory. Those studies determined the absence of any temperature-induced phase transition between the o-[Na 2 Cl]GaS 2 and t-[Na 2 Cl]GaS 2 polytypic compounds. Moreover, herein we probed the suitability of the heterolayered [Na 2 Cl]GaS 2 single crystals as a sorbent for UO 2 2+ uptake and monitored this process by energy-dispersive and infrared spectroscopies and PXRD, which revealed that the [Na 2 Cl]+ insert could be exchanged with UO 2 2+ on the surface while the UO 2 2+ intercalation decomposes the [Na 2 Cl]GaS 2 structure.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Infrared-active phonon modes in single-crystal thorium dioxide and uranium dioxide

The infrared-active phonon modes, in single-crystal samples of thorium dioxide (ThO 2 ) and uranium dioxide (UO 2 ), were investigated using spectroscopic ellipsometry and compared with density functional theory. Both ThO 2 and UO 2 are found to have one infrared-active phonon mode pair [consisting of one transverse optic (TO) and one associated longitudinal optic (LO) mode], which is responsible for the dominant features in the ellipsometric data. Furthermore, at room temperature, our results for the mode pair’s resonant frequencies and broadening parameters are comparable with previous reflectance spectroscopy characterizations and density functional theory predictions. For ThO 2 , our ellipsometry and density function theory results both show that the LO mode broadening parameter is larger than the TO mode broadening. This signifies mode anharmonicity, which can be attributed to the intrinsic phonon–phonon interaction. In addition to the main mode pair, a broad low-amplitude impurity-like vibrational mode pair is detected within the reststrahlen band for both ThO 2 and UO 2 . Elevated temperature measurements were performed for ThO 2 in order to study the mechanisms by which the phonon parameters evolve with increased heat. The observed change in the TO resonant frequency is in excellent agreement with previous density functional calculations, which only consider volume expansion of the crystal lattice. This suggests that the temperature-dependent change in the TO frequency is primarily due to volume expansion. The change in the main mode pair’s broadening parameters is nearly linear within the temperature range of this study, which indicates the intrinsic anharmonic scattering (via cubic anharmonicities) as the main decay mechanism.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Interaction of hydrogen with actinide dioxide (011) surfaces

The corrosion and oxidation of actinide metals, leading to the formation of metal-oxide surface layers with the catalytic evolution of hydrogen, impacts the management of nuclear materials. In this work, the interaction of hydrogen with actinide dioxide (AnO 2 , An = U, Np, or Pu) (011) surfaces by Hubbard corrected density functional theory (PBEsol+U) has been studied, including spin–orbit interactions and non-collinear 3k anti-ferromagnetic behavior. The actinide dioxides crystalize in the fluorite-type structure, and although the (111) surface dominates the crystal morphology, the (011) surface energetics may lead to more significant interaction with hydrogen. The dissociative adsorption of hydrogen on the UO 2 (0.44 eV), NpO 2 (-0.47 eV), and PuO 2 (-1.71 eV) (011) surfaces has been calculated. It is found that hydrogen dissociates on the PuO 2 (011) surface; however, UO 2 (011) and NpO 2 (011) surfaces are relatively inert. Recombination of hydrogen ions is likely to occur on the UO 2 (011) and NpO 2 (011) surfaces, whereas hydroxide formation is shown to occur on the PuO 2 (011) surface, which distorts the surface structure.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Chemical and elemental mapping of spent nuclear fuel sections by soft X-ray spectromicroscopy

Soft X-ray spectromicroscopy at the O K -edge, U N 4,5 -edges and Ce M 4,5 -edges has been performed on focused ion beam sections of spent nuclear fuel for the first time, yielding chemical information on the sub-micrometer scale. To analyze these data, a modification to non-negative matrix factorization (NMF) was developed, in which the data are no longer required to be non-negative, but the non-negativity of the spectral components and fit coefficients is largely preserved. The modified NMF method was utilized at the O K -edge to distinguish between two components, one present in the bulk of the sample similar to UO 2 and one present at the interface of the sample which is a hyperstoichiometric UO 2+ x species. The species maps are consistent with a model of a thin layer of UO 2+ x over the entire sample, which is likely explained by oxidation after focused ion beam (FIB) sectioning. In addition to the uranium oxide bulk of the sample, Ce measurements were also performed to investigate the oxidation state of that fission product, which is the subject of considerable interest. Analysis of the Ce spectra shows that Ce is in a predominantly trivalent state, with a possible contribution from tetravalent Ce. Atom probe analysis was performed to provide confirmation of the presence and localization of Ce in the spent fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

NOvA 2020 official data release (13.6E20 neutrino + 12.5E20 antineutrino)

NOvA results of the 2020 joint analysis of nue appearance and numu disappearance data presented in Phys. Rev. D 106, 032004:https://doi.org/10.1103/PhysRevD.106.032004Note: these contained plots are slightly different from the ones shown at NEUTRINO 2020.Exposure: 13.6E20 protons on target, neutrino-enhanced beam12.5E20 protons on target, antineutrino-enhanced beamParameters not constrained by NOvA:- th13 is assumed to be Gaussian, ss2th13 = 0.085 +/- 0.003 (PDG 2019, Phys. Rev. D98, 030001 (2018) and 2019 update, https://pdg.lbl.gov/2019/listings/rpp2019-list-neutrino-mixing.pdf)- ss2th12 = 0.851, dm21=7.53e-5 are fixed at 2019 PDG valuesSignificances are obtained with the Feldman-Cousins procedure.Files contain TGraphs for the following:- ssth23 vs deltaCP 1,2,3 sigma and 90% CL contours, for NO or IO- dmsq32 vs ssth23 1,2,3 sigma and 90% CL contours, for NO or IO- deltaCP profile for NO UO, NO LO, IO LO, IO UO- ssth23 profile for NO, IO- dmsq32 profile for NO UO, NO LO, IO LO, IO UOContour files also include a TMarker with the best-fit parameters (in the NO)An additional file containing the predictions for all channels’ signal and background components computed at the NOvA best-fit oscillation parameters and systematic pull terms. Numu samples include the no-oscillation case as well.Please note that any fits performed with these histograms are not expected to exactly reproduce the official NOvA results as parameterizations of the numerous systematic uncertainties considered in the official fits are not included in this release.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2021

Proton irradiation-induced blistering in UO2

Abstract Proton (H + ) irradiation effects in polycrystalline UO 2 have been studied. The irradiation was carried out using three ion energies and two different ion fluxes at 600 °C. Scanning electron microscopy (SEM) investigations showed that significant surface flaking took place. Focused ion beam (FIB) milling in SEM was successfully applied for extracting lamellas from uneven blistered surfaces for transmission electron microscopy (TEM) investigations allowing detailed investigations for the degradation mechanisms. High-resolution TEM for the flaked UO 2 surfaces revealed that the implanted H + formed sharp two-dimensional cavities at the peak ion-stopping region instead of diffusing to the matrix. The resulting lateral stress likely caused UO 2 surface deterioration in good agreement with previous blistering and flaking studies on crystalline materials. Graphical abstract

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

A Coupled Experimental and Simulation Approach to Investigate the Impact of Grain Growth, Amorphization, and Grain Subdivision in Accident Tolerant U 3 Si 2 Light Water Reactor Fuel

Triuranium disilicide (U 3 Si 2 ) is being considered as an alternative accident tolerant fuel (ATF) due to its higher thermal conductivity. However, there is uncertainty in its use due to a lack of irradiation data at light water reactor (LWR) conditions. In this project, we used an integrated experimental and simulation approach to answer two significant questions pertaining to U 3 Si 2 that were not investigated by the Fuel Cycle Research and Development (FCRD) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) programs: (1) Will grain growth in the hotter portions of the fuel significantly impact U 3 Si 2 LWR fuel behavior? (2) Under what conditions do grain subdivision and amorphization occur in U 3 Si 2 and will either occur at LWR conditions? Our approach to answer these questions used in situ ion irradiation and annealing to investigate grain growth, grain subdivision, and amorphization, along with mesoscale simulations using the MARMOT tool. We found that while the grain boundary mobility of U 3 Si 2 is larger than that of UO 2 , grain growth in U 3 Si 2 fuel pellets will be less than in UO 2 and may be neglected due to the lower thermal conductivity and thus lower centerline temperature. We also found that amorphization will not occur in U 3 Si 2 above 600 K, such that it is not likely to occur at typical LWR operating conditions. Grain subdivision will occur at LWR conditions and will likely occur at lower burnups than in UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Formation of Aerosol Nanoparticles by Gas-Phase Hydrolysis Reaction of Uranium Hexafluoride

The aerosol physics of uranyl particle formation has been addressed in this research using advanced aerosol instrumentation and an aerosol dynamics model. Based on the research works, we conclude that the formation and growth of aerosol particles by gas-phase UF6 hydrolysis strongly depends on the availability of water molecules in our reactor conditions. The total number concentration of the UO 2 F 2 particulate material that could be produced in the hydrolysis reaction is also regulated primarily by the availability of water molecule concentration. The higher the water molecule concentration, the higher the number and the larger the size of UO 2 F 2 aerosol particles that could be produced in a reactor custom-built at ORNL. Although the aerosol reactor was enabling the study of particle formation kinetics, the instrumentation was still insufficient in characterizing the chemical composition of the produced particles as well as the time-dependent evolution of the particulate species. The temporal evolution could impact the eventual fate of the particles upon release to the environment (i.e., the physio-chemical transformation, transport, and removal). On uranyl particle formation kinetics, we found that the growth rates of aerosol particles appeared to approach a single number in the range of 0.05 ± 0.03 - 0.08 ± 0.04 nm/s, statistically, as the ω value becomes smaller than 1. The size of primary particles from the UF6 hydrolysis at water-deprived condition was estimated to be 3.6 ± 0.4 nm; the higher the availability of water molecules, the larger the primary particles. The ability to precisely control the availability of water molecules in the reaction could lead to the production of nearly monodispersed aerosol particles. In other words, the result suggests that one can precisely manipulate the size of UO 2 F 2 aerosol particles by controlling the water vapor availability and interaction of water molecules with U F6 in the reaction. This finding has significant implications in the engineering manufacturing of fuel powder materials and possibly to future development and deployment of an environmental sampling apparatus.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2021

Calibration of the Diffusivity Predictions of Centipede Using Approximate Bayesian Computation and Applications in Nyx (Engineering Scale) and Xolotl-MARMOT (Meso-Scale) Simulations

Fission gas evolution and release in UO 2 nuclear fuel are important fuel performance metrics and occur in several distinct stages: 1) nucleation, growth and resolution of intra-granular bubbles, 2) diffusion to grain boundaries and 3) nucleation and growth of bubbles at grain boundaries, which eventually form a connected network (percolation) enabling release of gas from grain boundaries through connections to triple junctions, grain edges or free surfaces. The NE-SciDAC project is developing several computational tools to model this problem, which are connected in a hierarchical multi-scale framework. The information transfer in the multi-scale framework is a critical step that, in addition to best-estimates, should include uncertainty quantification. Despite taking a first-principles multi-scale approach, there is a need to perform parameter calibration to ensure consistency with available experimental data. In the present study, uncertainty quantification (UQ) and parameter calibration is demonstrated for one of the lower length scale codes in the multi-scale framework (Centipede) and then the results, including instances of the propagated uncertainties, are used in other codes within the framework, specifically Nyx and Xolotl-MARMOT. We calibrated the model parameters in Centipede, a computer code used to predict diffusivities of uranium (U) and xenon (Xe) in the context of the simulation of fission gas in uranium oxide (UO 2 ) nuclear fuel. The Centipede code depends on 183 parameters, all of which are subject to uncertainty. The three data sets used in our calibration effort are taken from the literature. This data is available as a set of measurements, including measurement errors. Our goal is to calibrate a statistical model that predicts both the value of the measurement and the uncertainty associated with the measurement. We perform a Bayesian calibration of the model parameters using a dedicated approximate Bayesian computation (ABC) likelihood function. To avoid excessive computational costs, we replace the expensive Centipede simulation code by a higher-order surrogate model, constructed using only the 9 most important parameters. These important parameters are identified by a preliminary global sensitivity analysis (GSA) study. Among the important parameters are T0 (the temperature at which UO 2 is perfectly stoichiometric) and Hf_pO2 (the temperature dependence of the oxygen (O) partial pressure) that should be considered as operating conditions to be estimated along with the other parameters. We consider two different cases: one where we define one set of these operating conditions for all data sets, and one where we define distinct operating condition parameters for each data set. The Xe diffusivities predicted by the latter case show distinct features that could not be observed in the former. Next, we use the diffusivity predictions by Centipede as input to Nyx, a reduced order fuel performance code focused on gas behavior alone, in order to estimate quantities associated with inter-granular bubble formation at conditions specified by the experiments. Finally, the diffusivities obtained from the calibrated Centipede runs were used in coupled Xolotl-MARMOT simulations of intra- and inter-granular gas evolution. The results are compared to simulations using the baseline diffusivities from Turnbull et al.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 1

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore fuel product. The long-term EnCore ® fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 2

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore ® fuel product. The long-term EnCore fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15 N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

NCSP Integral Experiments at Sandia in FY21 [Slides]

This presentation provides details regarding integral experiments at Sandia National Laboratory for fiscal year 2021. The experiments discussed are as follows: IER 230: Characterize the Thermal Capabilities of the 7uPCX; IER 304: Temperature Dependent Critical Benchmarks; IER 305: Critical Experiments with UO 2 Rods and Molybdenum Foils; IER 306: Critical Experiments with UO 2 Rods and Rhodium Foils ; IER 441: Epithermal HEX Lattices with SNL 7uPCX Fuel for Testing Nuclear Data; IER 452: Inversion Point of the Isothermal Reactivity Coefficient; and IER 523: Critical Experiments with ACRR UO 2 -BeO Fuel.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗