Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “U”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Thermal Conductivity Degradation in High Burnup U-Pu-Zr Fuel

Recent advancements in the characterization of irradiated U-Pu-Zr fuels have revealed complexities that challenge existing understanding of constituent redistribution. Traditionally, models have proposed three concentric regions within the fuel, each characterized by distinctive phases and porosity. However, through detailed analysis of high burnup U-Pu-Zr, we discovered the presence of four distinct constituent redistribution regions. Particularly novel is the observation of significant Pu redistribution, a previously unreported phenomenon that necessitates a reevaluation of current models. This work aims to delve deeper into these findings, seeking to correlate mesoscale measurements of thermal diffusivity and respective thermal conductivity with the phases present in each redistribution region. To achieve this objective, we employed mesoscale thermoreflectance methods using the unique, Idaho National Laboratory (INL) developed, Thermal Conductivity Microscope (TCM) at INL’s Irradiated Materials Characterization Laboratory. The TCM employs two tightly focused lasers: one for heating to generate periodic thermal waves in the substrate, and another spatially separated probe laser to detect changes in the optical reflectivity of the gold-coated substrate resulting from thermal wave diffusion. We conducted several thermal diffusivity measurements within each region of constituent redistribution of a U-19Pu-10Zr fuel pin cross section irradiated to 11 at. % burnup. The TCM measurement positions strategically aligned with transmission electron microscopy (TEM) lift-out locations previously collected from the fuel sample. Complementary microstructural analysis techniques such as optical and scanning electron microscopy (OM/SEM), electron probe microanalysis for chemical compositions, and TEM-based selective area electron diffraction (SAED) analysis for crystallographic insights into each phase were also utilized. This comprehensive approach allowed us to correlate local thermal diffusivity data with microstructural characteristics, enabling the computation of local thermal conductivity at each position. The significance of this contribution lies in its pioneering use of the TCM for ternary fuel mesoscale examination, shedding light on the previously overlooked effects of Pu redistribution on local thermal conductivity. By informing current models capturing constituent redistribution and heat transfer, our findings pave the way for more accurate predictions of metallic fuel performance. Moreover, this work sets the stage for future comparisons with similar TCM examinations on U-19Pu-10Zr fuels at ultra-low burnup, facilitating a comprehensive understanding of thermal property changes across different burnup levels. Ultimately, our study not only enriches our understanding of the thermophysical properties of individual redistribution regions within U-Pu-Zr fuel but also offers valuable insights for the design and operational parameters of proposed next-generation fast reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

DISSOLUTION OF SURROGATE U-Zr FUEL USING ALNIFLEX CONDITIONS

Non-aluminum clad spent nuclear fuel (NASNF) stored in L Basin at the Savannah River Site (SRS) is widely varied in fuel composition, design, packaging, and physical condition. The complexity of the NASNF inventory presents significant challenges, and technology development is necessary for successful disposition. One such fuel in the inventory is metallic uranium-zirconium (U-Zr) alloy fuel, the focus of this study. Electrolytic or nitric acid only dissolution of metallic U-Zr alloy can form insoluble zirconium oxide, which results in up to 52% loss of U to insoluble solids, and can be subject to potentially uncontrolled oxidation reactions [1, 2]. The AlNiflex process was determined to be a viable dissolution flowsheet for the U-Zr alloy fuel. Under a narrow set of solution concentrations, a combination of hydrofluoric acid (HF), nitric acid (HNO3), aluminum nitrate (Al(NO3)3), and hexavalent chromium can safely dissolve U-Zr intermetallic alloys, keep Zr soluble, and not significantly corrode stainless steel (SS) vessels [3, 4.

Gogolski, Jarrod M. [Savannah River National Labor↗
DOE OSTI2022

Understanding Fission Gas Bubble Distribution and Zirconium Redistribution in Neutron-irradiated U-Zr Metallic Fuel Using Machine Learning

U-10wt.% Zr (U-10Zr) based metallic fuel is the leading candidate for next-generation sodium cooled fast reactor in United States. Currently, Idaho National Laboratory (INL) has been the leading national laboratory for research, development, and demonstration (RD&D) on metallic fuel. Advanced post-irradiation characterization will help to understand fuel microstructure and property change during irradiation, benefiting fuel qualification for commercial application. Characterization capabilities ranging from sub-nanometer to micrometer, such as scanning electron microscopy (SEM), focused ion beam (FIB) sampling, transmission electron microscopy (TEM) characterization, and local thermal conductivity microscopy (TCM), have been utilized recently on irradiated U-10Zr fuel samples to gain a better understanding of nuclear fuel microstructure and property evolution inside a reactor. The FIB/SEM coupled with energy dispersive X-ray spectroscopy (EDS) can capture the essential information to achieve better understanding of fuel behaviors. Inside a nuclear reactor, the phase and microstructure of U-10Zr is constantly changing under neutron bombardment. For example, the gaseous fission product atoms have a limited solubility inside fuel matrix and tend to precipitate out in bubble form, which not only contribute to fuel thermal conductivity degradation but also provide a shortcut for movement of fission products, i.e. lanthanides. The resultant deposition of lanthanides at the cladding inner surface will potentially trigger a chemical reaction/interaction between nuclear fuel and cladding at reactor operational conditions, threatening fuel integrity and safety. FIB/SEM coupled with EDS can provide the fission bubble information as well as probe into phase separation or Zr redistribution, which is fundamental to predict the fuel performance. With high velocity image data generating method, such as FIB/SEM, an automatic way to extract the microstructural information quantitively can better serve the needs from post irradiation characterization. A trained machine learning model, named Decision Tree, is employed to generate a bubble classifier and to categorize bubbles into three categories: isolated bubble, connected without lanthanides, and connected with lanthanides bubbles[3]. This work presents a showcase of this approach on six regions of a fuel cross-section along the radial temperature gradient. We obtained distributions of bubble categories and porosity rates along the six regions. Moreover, a secondary phase U-Zr2 was determined and found on regions 5 and 6. The secondary phase fraction was increasing from 15.61% in region 5 to 34.79% in region 6 based on this approach . This quantitative data offers insights into the lanthanide migration and potentially thermal conductivity degradation. This information from machine learning will be fed into fuel design code for better prediction of fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A critical analysis of U-Pu-Zr phase transitions using calorimetric, microstructural, and phase equilibria data

Metallic fuels consisting primarily of uranium, plutonium, and zirconium (U-Pu-Zr) are a leading material candidate for fast-spectrum nuclear reactors. Early demonstration programs proved the principle of safe and efficient fast reactor operation, however there is still considerable uncertainty regarding the phase equilibria and microstructural evolution across the ternary composition space. Quantitative phase formation and identification measurements are scarce and often incomplete, with studies reporting either phase transition temperatures or phase identification data, but not both from the same specimens. In this study, we critically compared experimental and calculated phase transition data and correlated with the microstructure and phase characterization data of as-cast and annealed U-Pu-Zr alloys. Differential scanning calorimetry (DSC) was used to measure phase transitions in the subsolidus regions (723−948 K) of three ternary U-Pu-Zr alloys with similar plutonium concentrations but various U/Zr ratios. Due to sluggish kinetics and narrow ranges of phase stability, complex peaks required the use of a Frazier-Suzuki peak fitting algorithm to deconvolute and calculate transition peak temperatures and enthalpies. We also identified trends of phase transition behavior by critically comparing our DSC data with previous phase transition measurements as well as historical and calculated phase equilibrium diagrams. In conclusion, this provides a critical approach for benchmarking and assessing the quality of new U-Pu-Zr phase equilibria data prior to its incorporation into nuclear material databases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of ZrN coating and heat treatment on U-Mo dispersion fuel systems under irradiation

The stability of U-Mo fuel particles embedded in an Al matrix under irradiation can be enhanced through ZrN coatings and/or heat treatment. Here, the present study investigates the irradiation behavior of fuel plates containing U-Mo fuel particles fabricated under various heat-treatment conditions and ZrN coating thicknesses. Different fission densities were also applied to each fuel plate to evaluate the effects of these variables. Results indicate that higher fission densities lead to more grain recrystallization and high burnup structure (HBS) development in the fuel particles. Heat treatment was found to mitigate the accumulation of fission gas bubbles in fuel particles at low fission densities by coarsening their grains. Fuel particles with ZrN coatings of a 1.2 μm thickness or above exhibited reduced formation of U-Mo/Al interaction layers, suggesting the existence of a critical ZrN coating thickness that minimizes the development of these layers. Fission gas bubbles were predominantly observed at grain boundaries of U-Mo fuel particles irradiated at low fission densities. Subgrain boundaries, which appeared to originate from the original grain boundaries containing fission gas bubbles or HBSs, were also observed, indicating the early stage of HBS propagation in the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Materials Data on U(Al10V)2 by Materials Project

UV2Al20 crystallizes in the cubic Fd-3m space group. The structure is three-dimensional. U is bonded in a 4-coordinate geometry to sixteen Al atoms. There are four shorter (3.13 Å) and twelve longer (3.17 Å) U–Al bond lengths. V is bonded to twelve Al atoms to form VAl12 cuboctahedra that share corners with six equivalent VAl12 cuboctahedra, edges with eighteen equivalent AlUAl10V cuboctahedra, and faces with six equivalent AlUAl10V cuboctahedra. There are six shorter (2.56 Å) and six longer (2.82 Å) V–Al bond lengths. There are three inequivalent Al sites. In the first Al site, Al is bonded in a distorted linear geometry to two equivalent U and twelve equivalent Al atoms. All Al–Al bond lengths are 3.07 Å. In the second Al site, Al is bonded in a 12-coordinate geometry to two equivalent V and ten Al atoms. There are a spread of Al–Al bond distances ranging from 2.72–2.85 Å. In the third Al site, Al is bonded to one U, one V, and ten Al atoms to form distorted AlUAl10V cuboctahedra that share corners with fifteen equivalent AlUAl10V cuboctahedra, edges with two equivalent AlUAl10V cuboctahedra, edges with three equivalent VAl12 cuboctahedra, a faceface with one VAl12 cuboctahedra, and faces with fifteen equivalent AlUAl10V cuboctahedra. There are a spread of Al–Al bond distances ranging from 2.67–2.93 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on U(Zn10Os)2 by Materials Project

UOs2Zn20 crystallizes in the cubic Fd-3m space group. The structure is three-dimensional. U is bonded in a 4-coordinate geometry to sixteen Zn atoms. There are four shorter (3.10 Å) and twelve longer (3.13 Å) U–Zn bond lengths. Os is bonded to twelve Zn atoms to form OsZn12 cuboctahedra that share corners with six equivalent OsZn12 cuboctahedra, edges with eighteen equivalent ZnUZn10Os cuboctahedra, and faces with six equivalent ZnUZn10Os cuboctahedra. There are six shorter (2.54 Å) and six longer (2.81 Å) Os–Zn bond lengths. There are three inequivalent Zn sites. In the first Zn site, Zn is bonded in a 2-coordinate geometry to two equivalent U and twelve equivalent Zn atoms. All Zn–Zn bond lengths are 3.04 Å. In the second Zn site, Zn is bonded in a distorted linear geometry to two equivalent Os and six equivalent Zn atoms. There are two shorter (2.69 Å) and four longer (2.83 Å) Zn–Zn bond lengths. In the third Zn site, Zn is bonded to one U, one Os, and ten Zn atoms to form distorted ZnUZn10Os cuboctahedra that share corners with fifteen equivalent ZnUZn10Os cuboctahedra, edges with two equivalent ZnUZn10Os cuboctahedra, edges with three equivalent OsZn12 cuboctahedra, a faceface with one OsZn12 cuboctahedra, and faces with fifteen equivalent ZnUZn10Os cuboctahedra. There are a spread of Zn–Zn bond distances ranging from 2.64–2.92 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on U(Ge3Pt)4 by Materials Project

U(PtGe3)4 crystallizes in the cubic Im-3 space group. The structure is three-dimensional. U is bonded to twelve equivalent Ge atoms to form UGe12 cuboctahedra that share faces with eight equivalent PtGe6 octahedra. All U–Ge bond lengths are 3.33 Å. Pt is bonded to six equivalent Ge atoms to form PtGe6 octahedra that share corners with six equivalent PtGe6 octahedra and faces with two equivalent UGe12 cuboctahedra. The corner-sharing octahedral tilt angles are 59°. All Pt–Ge bond lengths are 2.51 Å. Ge is bonded in a 2-coordinate geometry to one U, two equivalent Pt, and two equivalent Ge atoms. There are one shorter (2.58 Å) and one longer (2.62 Å) Ge–Ge bond lengths.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Effects of ZrN Coating and Heat Treatment on U-Mo Dispersion Fuel under Irradiation

The stability of U-Mo fuel particles in Al matrix can be improved using ZrN coating and heat treatment. This study investigated irradiation behaviors of fuel plates containing U-Mo fuel particles fabricated with different conditions including heat treatment and the thickness of ZrN coating. Additionally, different fission densities were applied to each fuel plate to understand how varying fission densities affect its irradiation behaviors. This study showed that more recrystallization of grains and development of high burnup structure (HBS) occurred in irradiated fuel particles with higher fission densities. Heat treatment reduced the accumulation of fission gas bubbles in the fuel particles by coarsening their grains. The fuel particles with a ZrN coating thickness of 1.2-1.8 µm showed less significant development of U-Mo/Al interaction layers than the particles with a coating thickness of 0.5 µm after irradiation. This might indicate the existence of a critical thickness of ZrN coating to minimize the formation of the interaction layers. Fission gas bubbles were observed at grain boundaries of irradiated U-Mo fuel particles at low fission densities, and the formation of dendritic features with a darker contrast was identified in backscattered electron images. These features stemmed from grain boundaries with fission gas bubbles or HBS, and some of them contained the bubbles, indicating that they might represent an early stage of HBS propagation in the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Investigation of Drift Effects in UEDGE Simulations of NSTX-U Edge Plasma With Lithium Divertors

Lithium is a low-Z material, and lithium-based plasma-facing components (PFCs) are planned for the National Spherical Torus Experiment Upgrade (NSTX-U) to explore potential benefits for divertor power exhaust and core plasma management. NSTX-U is a medium-sized spherical tokamak with up to 12 MW of auxiliary heating, capable of generating reactor-relevant plasma conditions. This work presents boundary plasma simulations for NSTX-U with lithium PFCs using the UEDGE code, incorporating full magnetic and 𝐄 ×𝐁 drift physics. The simulations show that drifts strongly influence heat and particle transport: they enhance convective transport, broaden the scrape-off layer heat-flux width 𝜆 𝑞 , and reduce the anomalous heat diffusivity 𝜒 required to reproduce predicted SOL heat-flux width. 𝐄 ×𝐁 drifts provide poloidal transport, while ∇𝐵 (which includes both gradB and curvature) drifts provide radial heat and particle transport. Lithium transport is also affected by drifts, with lithium ions migrating from the outer divertor to the inner divertor through the private flux region (PFR) following the 𝐄 ×𝐁 drifts flow, lowering upstream impurity lithium densities. UEDGE is self-consistently coupled with the Wall-Li model to study plasma lithium PFC interactions depending on the local lithium sourcing based on local plasma conditions and lithium surface temperature. In these simulations, lithium evaporation shows a vapor-shielding effect that reduces divertor heat flux and increases radiative losses once surface temperatures exceed 450°C. This research work provides a first step toward self-consistent modeling of lithium PFCs in NSTX-U, demonstrating the impact of drift-driven plasma transport in SOL and divertor regions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Design of Mini-Plate-1 Irradiation Test for Qualification of High-Density, Low-Enriched U-10Mo Monolithic Fuel

The United States High Performance Research Reactor project is tasked with fuel development and qualification leading to conversion of higher power research and test reactors in the US from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuels. Here, this manuscript identifies the functional and operational design requirements of the first miniature test plate (mini-plate [MP]) irradiation campaign (MP-1) of commercially fabricated LEU U-10Mo monolithic plate-type fuel and is the precursor to a large parametric mini-plate test (MP-2) aimed at producing the data to support regulatory qualification of the LEU U-10Mo monolithic fuel. The manuscript (1) provides a general description of the selected U-10Mo LEU fuel and (2) defines the overall experiment design and functional requirements to accomplish the specific test objective of MP-1, which is to confirm that the commercially manufactured LEU U-10Mo monolithic fuel meets the established requirements of geometric stability, mechanical integrity and stable and predictable behavior. The fuel testing parameters are established by the need to bound performance behavior within the operational envelope of the reactors being converted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Integrating Analytical Solutions and U-Net Model for Predicting Groundwater Contaminant Plumes in Pump-and-Treat Systems

Pump-and-treat (P&T) is a common technique for groundwater remediation involving the extraction and treatment of contaminated water above ground. Optimizing the design and operation of the P&T well network is essential for maximizing the system’s effectiveness and efficiency. However, this optimization often necessitates many model evaluations, leading to computationally demanding tasks. This study introduces a novel approach that integrates analytical solutions for groundwater dynamics with the U-Net (Ronneberger et al., 2015) deep learning framework to predict groundwater contaminant plume migration under dynamic pumping conditions. By incorporating the Thiem equation (Thiem, 1906) into the input preprocessing, the U-Net model transforms sparse well data into a continuous spatial field that captures the hydraulic impacts of pumping activities. This integration enables the model to leverage both deep learning capabilities and classical physics-based groundwater theories, enhancing prediction accuracy and computational efficiency. These advancements can facilitate rapid, large-scale evaluations of P&T optimization simulations, allowing for timely and effective decision-making in well placement and system management. We demonstrate the model's robust performance across both simplified transient 2D models and a more complex 3D heterogeneous site model at the 200 West P&T facility at the Hanford Site. The U-Net-based model offers substantial computational advantages, reducing simulation times significantly compared to full physics-based models and providing a powerful tool for rapid site evaluation and P&T system optimization, such as evaluating alternative P&T well network designs. Our findings highlight the potential of advanced machine learning models to significantly enhance the efficiency and sustainability of groundwater remediation efforts, offering a novel application of U-Net architecture in environmental science.

Pump-and-treat↗
DOE OSTI2021

Progress on the reevaluation and validation of the n+233U neutron cross sections

The set of 233 U resonance parameters of the ENDF/B-VIII.0 nuclear data library was adopted from the previous ENDF/B-VII.1 evaluation using the external levels to update the thermal values. Adoption of IAEA 2017 thermal standards ( σ f = 533.0 ± 2.2 b, σ c = 44.9 ± 0.9 b, and ν ‾ tot = 2.487 ± 0.011 ) and of the IAEA-recommended thermal-neutron induced prompt fission neutron spectrum (PFNS) with average PFNS energy of 2.030 ± 0.013 MeV requires a re-evaluation of 233 U neutron cross sections in the resolved resonance region. A newly produced evaluation is being tested on benchmarks carefully selected from the Handbook of International Criticality Safety Benchmark Experiments (ICSBEP) which are highly sensitive to 233 U data. An important goal of this work was to eliminate the strong negative gradient of the calculated effective multiplication factors with respect to the epithermal fission fraction observed in the validation of the ENDF/B-VIII.0 library for those assemblies. A significant improvement in integral performance of critical 233 U solutions is observed for the newly proposed evaluation. Further work addressing the fast neutron region is needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Improving stability of ALD ZrN thin film coatings over U-Mo dispersion fuel

Atomic layer deposition (ALD) of ZrN is a candidate technology for coating U-Mo dispersion fuel as a diffusion barrier coating. During the early development stages of the coating the ZrN deposition showed to be mechanically unstable and ultimately resulted in spalling. Based on experimental outcomes it was found that instabillity can be eliminated through the introduction of a thin amorphous (a) Al 2 O 3 interlayer coating deposited in between the ~1-μm thick ZrN and the U-Mo. substrate. To elucidate these findings simulations were performed with density functional theory (DFT) to measure work of adhesions at different interfaces while finite element modeling (FEM) was performed to measure the residual stress distribution. DFT indicated that the ZrN coating deposited over (a)-Al 2 O 3 is ~2.5 times stronger when compared to direct deposition over U or UO 2 substrate. While calculations from FEM recognized; (1) large stress concentrations can originate from the irregularly distributed native surface oxides (UO 2 ) over U-Mo and (2) stress concentrations can be reduced if those surface UO 2 can be modified to a uniform layer (achieved after application of a 8 nm (a)-Al 2 O 3 interlayer). These results provided explanations and confirmed the role played by interlayer in eliminating the original mechanical instability of the ZrN.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Self-consistent equilibrium and transport simulations for NSTX-U plasmas enhanced via machine learning surrogate models

The Control-Oriented Transport SIMulator (COTSIM) is an advanced equilibrium and transport code designed for simulating tokamak discharges at computational speeds suitable for control applications. COTSIM’s modular framework enables users to select models that balance accuracy with speed according to specific needs, allowing the code to operate from fast to faster-than-real-time performance levels. This work presents recent enhancements to COTSIM’s predictive accuracy for NSTX-U scenarios, achieved by integrating neural-network-based surrogate models and self-consistent equilibrium calculations. To improve source deposition predictions, a surrogate model for NUBEAM has been incorporated. Additionally, a surrogate model for the Multi-Mode Module (MMM) now supports predictions of anomalous thermal, momentum, and particle diffusivities—key factors for modeling the evolution of temperature and rotation. Each surrogate model was specifically trained for the NSTX-U operational regime to enhance COTSIM’s accuracy while maintaining computational efficiency. Moreover, COTSIM now couples fixed-boundary equilibrium solvers with its transport solvers, enabling self-consistent predictions of plasma profiles and equilibrium evolution over the discharge. Simulation results demonstrate strong agreement between COTSIM and TRANSP predictions for NSTX-U discharges. These substantial advancements expand COTSIM’s utility in model-based control applications for NSTX-U. Potential applications include simultaneous optimization of equilibrium and transport scenarios, integration into digital twins, real-time profile estimation (e.g., temperature and rotation) from limited or noisy measurements, and advanced feedback-based scenario control.

Equilibrium and transport modeling↗
DOE OSTI2020

Spark plasma sintering (SPS) densified U 3 Si 2 pellets: Microstructure control and enhanced mechanical and oxidation properties

Dense U 3 Si 2 pellets with controlled grain structure and enhanced thermal-mechanical and oxidation properties are synthesized with spark plasma sintering (SPS). Microstructure and phase composition of the SPS densified pellets are characterized systematically using SEM, EDS, and XRD. Thermal-mechanical properties and oxidation behavior of the sintered silicide fuel pellets are analyzed by laser flash, indentation, and dynamic thermogravimetric analysis. Dense U 3 Si 2 pellets are consolidated by combining high energy ball milling and rapid sintering by SPS, and the microstructure structures are controlled from micron-sized (~5.7 μm grain size) for conventional silicide to a nanocrystalline matrix with an average grain size of ~280 nm. A dominant phase of distorted U 3 Si 2 was identified with lattice expansion due to residual thermal stress upon SPS consolidation and rapid cooling processes. Both micron-sized and nano-sized pellets show exceptional thermal transport properties, consistent with monolithic silicides reported in literature. The SPS-densified pellets possess simultaneously high hardness and fracture toughness. The SPS-densified silicide pellets also demonstrate exceptional oxidation performance with extended onset oxidation temperature above 500 °C and reduced oxidation kinetics, particularly for nano-sized pellets. A strong strain effect was proposed in which compressive stress in nano-sized pellets enhances the oxidation resistance of silicide fuels, as evidenced by the degradation of oxidation performance upon strain relaxation by isothermal annealing. The correlation among the sintering process – microstructure control – physical properties and fuel behavior is established. A new concept of strain engineering is proposed further properties optimization, enabling the development of potential oxidation and corrosion-resistant silicides with extended performance, the key technological challenge of U 3 Si 2 as the leading concept of accident tolerant fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗